ML030920393

Proposed TS & Bases Amendment 2.0, Safety Limits; 3.3, Instrumentation; 3.4, Requirements

ML030920393
Person / Time
Site:	Mcguire, Catawba, McGuire   (NPF-009, NPF-017, NPF-035, NPF-052)
Issue date:	03/24/2003
From:	Mccollum W Duke Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML030920393 (59)

Text

{{#Wiki_filter:Duke PhPower. A DOd EVt CrpG y Duke Power 526 South Church Street PO. Box 1006 Charlotte, NC 28201-1006 March 24, 2003 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Duke Energy Corporation McGuire Nuclear Station, Units 1 and 2 Docket Numbers 50-369 and 50-370 Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413, 50-414 Proposed Technical Specifications and Bases Amendment 2.0, Safety Limits; 3.3, Instrumentation; 3.4, Reactor Coolant System; 5.6, Reporting Requirements In accordance with the provisions of 10CFR50.90, Duke Energy Corporation (Duke) proposes to revise the McGuire and Catawba Nuclear Station Facility Operating Licenses and Technical Specifications (TS) and Bases. The proposed amendment reduces the required minimum measured Reactor Coolant System (RCS) flow rate from 390,000 gallons per minute (gpm) to 382,000 gpm for McGuire and Catawba Units 1 and 2. Additionally, the proposed change relocates RCS related cycle-specific parameter limits from the TS to, and thus expands, the Core Operating Limits Reports (COLR) for the McGuire and Catawba Nuclear Stations. These proposed changes will allow Duke the flexibility of enhancing operating and core design margins without the need for cycle-specific license amendment requests. The proposed amendment changes TS 2.2.1, "Reactor Core Safety limits," and associated Bases, by relocating the reactor core safety limit figure to the COLRs and replacing it with specific fuel Departure from Nucleate Boiling Ratio (DNBR) and peak fuel centerline temperature safety limit requirements. The amendment proposes to change TS Table DDI

U.S. Nuclear Regulatory Commission Page 2 March 24, 2003 3.3.1-1, "Reactor Trip System Instrumentation," by relocating Overtemperature AT and Overpower AT nominal RCS operating pressure, nominal T.,,, and constant (K) values to the respective station COLR. The amendment also revises TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," and associated Bases, by relocating the pressurizer pressure, RCS average temperature, and RCS total flow rate values to the respective station COLR. The minimum RCS total flow rates are retained in TS 3.4.1.3. TS 5.6.5, "Core Operating Limits Report (COLR)," would be modified to reflect the above relocations to the COLR. The requested changes are based upon NRC approved Westinghouse Owners Group (WOG) Technical Specifications Task Force (TSTF) TSTF-339, "Relocate TS Parameters to the COLR Consistent with WCAP-14483," Revision 2, and Westinghouse WCAP-14483-A, "Generic Methodology for Expanding Core Operating Limits Report." The return to a lower RCS minimum flow rate, previously reviewed and approved by the NRC, is similar to that submitted by the Comanche Peak Nuclear Station on May 24, 1999, 7and approved by the NRC in the Safety Evaluation Report dated August 30, 1999. Duke requests approval of the proposed changes by August 1, 2003 to preclude the need to generate cycle-specific license amendment requests lowering the RCS total flow rates prior to the commencement of McGuire Unit 2, Cycle 16, and Catawba Unit 1, Cycle 15, operations. Duke has determined that the NRC's standard 30 day grace period will be sufficient for the implementation of this amendment. The contents of this amendment package are as follows: A provides marked copies of the affected TS and Bases pages for McGuire showing the proposed changes. B provides marked copies of the affected TS and Bases pages for Catawba showing the proposed changes. A, containing reprinted pages of the affected

U.S. Nuclear Regulatory Commission Page 3 March 24, 2003 TS and Bases pages for McGuire, will be provided to the NRC upon issuance of the approved amendment. Attachment 2B, containing reprinted pages of the affected TS and Bases pages for Catawba, will be provided to the NRC upon issuance of the approved amendment. provides a description of the proposed changes and technical justification. Pursuant to 10CFR50.92, Attachment 4 contains the results of the No Significant Hazards determination. Pursuant to 10CFR51.22(c)(9), Attachment 5 provides the basis for the categorical exclusion from the performance of an Environmental Assessment/Impact Review. Implementation of this amendment will impact the McGuire and Catawba Updated Final Safety Analysis Reports (UFSAR). Changes to the UFSARs will be submitted in accordance with 10CFR50.71(e) requirements. In accordance with Duke administrative procedures and Quality Assurance Program Topical Report requirements, this proposed amendment has previously been reviewed and approved by the McGuire and Catawba Plant Operations Review Committees and the Duke Corporate Nuclear Safety Review Board. Pursuant to 10CFR50.91, a copy of this proposed amendment is being sent to the appropriate state officials. Inquiries on this matter should be directed to J. A. Effinger at (704) 382-8688. Very truly yours, W. R. Mc Collum Senior Vice President Nuclear Support

U.S. Nuclear Regulatory Commission Page 4 March 24, 2003 W. R. Mc Collum, Jr., being duly sworn, affirms that he is the person who subscribed his name to the foregoing statement, and that all matters and facts set forth herein are true and correct to the best of his knowledge. W. R. Mc Collum, Senior Vice President, Nuclear Support Subscribed and sworn to me: My commission expires: Date Notary Public AJR 22- -)00( SSEAL

U.S. Nuclear Regulatory Commission Page 5 March 24, 2003 xc (w/attachments): L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 R. E. Martin NRC Project Manager (MNS) (CNS) U. S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 S. M. Shaeffer Senior Resident Inspector (MNS) U. S. Nuclear Regulatory Commission McGuire Nuclear Site E. F. Guthrie Senior Resident Inspector (CNS) U. S. Nuclear Regulatory Commission Catawba Nuclear Site B. 0. Hall, Section Chief Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645 R. Wingard, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

U.S. Nuclear Regulatory Commission Page 6 March 24, 2003 bxc w/attachments: C. J. Thomas G. D. Gilbert M. T. Cash G. B. Swindlehurst R. A. Hight L. J. Rudy K. E. Nicholson J. M. Ferguson (RGC data file) K. L. Crane R. L. Gill Catawba Owners: NCMPA-1, SREC, PMPA, NCEMC McGuire Master File (MG0lDM) Catawba Document Control File 801.01 (CN04DM) ELL

ATTACHMENT 1A McGUIRE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES MARKED COPY

INSERT 1 the COLR for four loop operation; and the following SLs shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 2 the 95/95 DNB criterion for the DNB correlation(s) specified in the thermal-hydraulic analytic method(s) listed in Section 5.6.5(b). 2.1.1.2 The peak fuel centerline temperature shall be maintained < the 95/95 CFM criterion for the fuel design(s) specified in the fuel mechanical analytic method(s) listed in Section 5.6.5(b). INSERT 2 The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95%

confidence level (the 95/95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and

b. There must be at least a 95% probability at a 95%

confidence level that the hot fuel pellet in the core does not experience centerline fuel melting. The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Over Temperature and Overpower LT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and AI that the reactor

core SLs will be satisfied during steady state operation, normal operational transients, and AO0s. INSERT 3 The numerical limits of these variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow, based on previously analyzed maximum steam generator tube plugging, is retained in the TS LCO.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the spcified inWW12,r 2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig. 2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. McGuire Units 1 and 2 2.0-1 Amendment Nos.,1IR¶~6

SLs 2.0 660 650 640 U-0 t3) (U Fn W/ 0l 630 DO NOT OPERATE IN THIS OPRTO 945 psia\\ .OPERATION\\ 620 610 600 590 580 0.2 0.4 0.6 0.8 1.0 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits - Four Loops in Operation McGuire Units I and 2 2.0-2 Amendment Nos. b Unit T Yi (Unit 2)

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7) Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.4% of RTP. kT (1+ TIs) (1 )*AT 0 IK1-K 2 (1+s [T 1 -Ti ÷K 3 (P-P)-fj 1 A)l (1 +-r2 5) (1 + r,3 5 S 2 + rs 5) IJ (1 + r6 5)- ] K,(-P -{ ^l Where: AT is measured RCS AT by loop narrow range RTDs, aF. ATo is the indicated AT at RTP, 'F. s is the Laplace transform operator, sec-1. T is the measured RCS average temperature, 'OF T is the nominal Tavg at RTP, <,{ F. _ = P is the measured pressurizer pressure, psig ) P is the nominal RCS operating pressure, =, L f A K1 = OvertemperatureAT reactor trip NOMINAL MP SETPOINT, as presented in the COLR, K2 = Overtemperature AT reactor trip heatup setpoint penalty coefficient, as presented in the COLR. K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR, x1.,2 = lime constants utilized in the leadlag controller for AT, as presented in the

COLR, T3

= Time constants utilized in the lag compensator for AT, as presented in the

COLR,

-4. Ts = Time constants utilized in the lead-lag controller for Tea, as presented in the

COLR,

-6 = Time constants utiftzed in the measured Tv lag compensator, as presented in the COLR, and, fl(AI) = a function of the Wiicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that (i) for q, - qb between the "positive and negativem f(Al) breakpoints as presented in the COLR; f1(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued) McGuire Units I and 2 3.3.1-18 Amendment Nos. J 5T i

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7) Reactor Trip System Instrumentation (ii) for each percent imbalance that the magnitude of qt - qb is more negative than the fI(Al) "negative' breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f I(Al) 'negative' slope presented in the COLR; and (iii) for each percent imbalance that the magnitude of qt - q, is more positive than the f (Al) 3positives breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f1(Al) "positive" slope presented in the COLR. Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 3.0%/* of RTP. AT (1 +TS ( 1 sa {5 ( 1 -T -Kr, 1 TT1 ( (A-(1 + ¶2 S t+ r3S JAOj(-S1.2r, st1+? 5T -K 6 [ 1+T sJj Where: AT is measured RCS AT by loop narrow range RTDs, 0F. ATo is the indicated AT at RTP, "F. s is the Laplace transform operator, seca'. T is the measured RCS average temperature 0 T is the nominal T.< at RTP, <,4, K4 = Overpower AT reactor NOMINAL TRIP SETPOINT as presented in the COLR,. Ks = Af for increas u r n rage temperature,a 1<

Overpower AT reactor trip heatup setpolnt penalty coefficient as presented intheCOLRforT>T'andK6e

forT at. 2= Time constants utilized In the l c oler for AT, as presented in the

COLR, T3

= Time constants utilized in the lag compensator for AT. as presented in the

COLR, x6

= Toine constants utilized in the measured Tl lag compensator, as presented in the COLR, v7

Time constant utilized in the rate-ag controller for T.., as presented in the COLR, and f2(AI)

a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (continued) McGuire Units 1 and 2 3.3.1-1 9 Amendment Nos.

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressureKRCS average temperature rhall be within th Thits specified g APPLICABILITY: MODT t-- Pressurizer pressure limit does not apply during:

a.

THERMAL POWER ramp > 5% RTP per minute; or

b.

THERMAL POWER step > 10% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours RCS average to within limit. temperature DNB parameters not within limits. B. RCS total flow rate / ) add- <1 eaJ Ad -i q /e O 9A Jirsecrg7 r ( /4 of 1Z B.1 Reduce THERMAL POWER to < 98% RTP. AND 2 hours 6 hours B.2 Reduce the Power Range Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP. ,I. (continued) McGuire Units 1 and 2 3.4.1 -1 Amendment Nos. tj4Qif ) )7z-Unut2-

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rate .4 q9 D 0' VA V& Xc

t,6-iE~k z

,-~ Coat.) 2:1VA,-, e/ 4 C.2.1 Reduce THERMAL POWER to < 50% RTP. AND C.2.2 Reduce the Power Range Neutron Flux - High Trip Setpoint to < 55% RTP. AND C.2.3 Restore RCS total flow rate 0 a X o Au X % e ) 2 hours 6 hours 24 hours I D. Required Action and D.1 Be in MODE 2. 6 hours associated Completion Time not met. McGuire Units 1 and 2 3.4.1 -2 Amendment Nos. 1 5 U ~ 17,7ic-2)

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 eure is within limits, 12 hours SR 3.4.1.2 Vperature is within limitg 12 hours S R 3.4.1.3 Verif RCS total flow rate is - AGED Eva -jo 0.(- 6zocG? NovA t /3,r 5pf co, -d nbe-, Caert 12 hours SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total flow indicator. 18 months I _____________________________________________________________ McGuire Units 1 and 2 3.4.1 -3 Amendment Nos. jI11

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1) RCS DNB Parameters 1. 4 3 < 587.2 'F < 586.9 'F Temperature 4 3 < 587.7 'F < 587.5 'F

2.

Indicated Pressurizer 4 3 > 2219.8 psig > 2222.1 psig Pressure 4 3 > 2215.8 psig > 2217.5 psig

3.

RCS Total Flow Rate > 390,000 gpm McGuire Units 1 and 2 3.4.1 -4 Amendment Nos. I

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A0Os. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.

There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b.

The hot fuel pellet in the core must not experience centerline fuel melting. The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature pressure, and THERMAL POWER level that would resul Itla eparture rom nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities. Automatic enforcement of these reactor core SLs is provided by the getIA11' A ? / g ( P A PC A.0 s-£ ,11. / tow piteseuAKpLuirriti\\ ,o / A~esT4rirS, J1*/ AdP6WdrAKT~fri' A*. /,P9ke/43pe)44kP(iqA r.41rd Ifr / AtaK~Kt~fe44yeV Iot I bees th or _ens es at

S esi a ameaste of The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

McGuire Units 1 and 2 B 2.1.1-2 Revision No-

Reactor Core SLs B 2.1.1 T S The 06 provided in XVj the loci of points of ATi4?' ce. -,J c-- SRCS Pressure, and average temperature for which the minimum DNBR is f~Argo0 -rve.m*c-e not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation. APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, 'Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER. SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 VIOLATIONS places the unit in a MODE in which this SL is not applicable. The allowed Completion Time of 1 hour recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage. REFERENCES

1.

10 CFR 50, Appendix A, GDC 10.

2.

UFSAR, Section 7.2. 3/ /p&y6AZ2/ wichysp a. IU1 McGuire Units 1 and 2 B 2.1.1 -3 Revision No. k

Reactor Core SLs B 2.1.1 rE tU-I N T (of) hVa OT&T £ORE Figure B 2.1.1-1 Illustration of Overtemperature and Overpower AT Protection McGuire Units 1 and 2 B 2.1.1-4 Revision No.,A

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure protection when below MODE 4.

9.

Pressurizer Water Level-High The Pressurizer Water Level-High trip Function provides a backup signal for the Pressurizer Pressure-High trip and also provides protection against water relief through the pressurizer safety valves. These valves are designed to pass steam in order to achieve their design energy removal rate. A reactor trip is actuated prior to the pressurizer becoming water solid. The setpoints are based on percent of instrument span. The LCO requires three channels of Pressurizer Water Level-High to be OPERABLE. The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is not required to address control/protection interaction concerns. The level channels do not actuate the safety valves, and the high pressure reactor trip is set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip. In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and take corrective actions.

10.

Reactor Coolant Flow-Low

a.

Reactor Coolant Flow-Low (Single Loon) The Reactor Coolant Flow-Low (Single Loop) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow. Above the P-8 setpoint, which is approximately 48% RTP, a loss of flow in any RCS loop will actuate a reactor trip. The setpoints are based on,%minimum R S flow. McGuire Units 1 and 2 B 3.3.1-17 Revision No,jQK

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) Cs-w, 0 W. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input. The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8. In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 1 O.b) because of the lower power level and the greater margin to the design limit DNBR.

b.

Reactor Coolant Flow-Low (Two Loops) The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow. Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The setpoints are based onhinimum o f low Q fJZ -J Co.4 I. Each loop has three flowdetectors to monitor flow. The flow signals are not used for any control system input. The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE. In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR. McGuire Units 1 and 2 B 3.3.1-1 8 Revision No.yf

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) assessed for their impact on the acceptance criteria. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, 'Control Bank Insertion Limits"; LCO 32.3, "AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, "QU BANT POWER TILT RATIO (QPTR).A The pressurizer pressure limits and the RCS average temperature limits correspond to analytical limits used in the safety analyses, with allowance for measurement uncertainty. The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2). LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses.1Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion. RCS total flow rate contains a measurement errord/;t/X based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner. Therefore, a penalty OWI4Afor undetected fouling of the feedwater venturi raises the nominal flow measurement allowance -'°/pfor no fouling. The numerical values)y&A ,0for pressure and average temperatur re given for the measurement location with adjustments for the indication instruments. APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern. McGuire Units 1 and 2 B 3.4.1-2 Revision No 1t

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued) A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the variations. l-i iprf~l~cX ris provided in SL 2.1.1, TAge.. C~tT~4~ ctr,. JJ IReactor Core SLs." ) 'are less restrictive than the limits of this 8- -'.r~. Gd LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded. ACTIONS A.1 Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s). The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience. B.1 and B.2 faEls at-'o'e RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is la wr.r. f Z. a w 'then THERMAL POWER may not 'ZQ o 4-exceed 98% RTP. THERMAL POWER must be reduced within 2 hours. The completion time of 2 hours is consistent with Required Action A.1. In 19 e / o

7) jLu addition, the Power Range Neutron Flux - High Trip Setpoint must be reduced from the nominal setpoint by 2% RTP within 6 hours. The Completion Time of 6 hours to reset the trip setpoints recognizes that, Ct a z R with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

McGuire Units 1 and 2 B 3.4.1-3 Revision No.;8

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued) I5sf 1J c4c gz -nK VALz '5T C a Z e,< Xg/' e jCw C.1, C.2.1, C.2.2, and C.2.3 If the indicated RCS total flow rate is W,0 4 then RCS total flow must be restore-dto h in,)f/p~ithin 2 hours or power must be ( ~ r~t? 7nJ oR \\reduced to less than 50% RTP. The Completion Time of 2 hours is _4 n consistent with Required Action A.1. If THERMAL POWER is reduced to P less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint iW~ V'4 2 Saftc-,F, must also be reduced to < 55% RTP. The Completion Time of 6 hours to (A*J Soc.-reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection System. Operation is permitted to continue provided the RCS total flow is restored to9 it ithin 24 hours. The Completion Time of 24 hours is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint. D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours. The Completion Time of 6 hours is reasonable to reach the required plant conditions in an orderly manner. SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. McGuire Units 1 and 2 B 3.4.1-4 Revision No.A\\

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radioloqical Environmental Operating Report (continued) The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of the analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 5.6.3 Radioactive Effluent Release Report -NOTED A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in Chapter 16 of the UFSAR and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1. 5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: -. 7X Moderator Temperature Coefficient BOL and EOL limits and 300 pmlveillance limit for Specification 3.1.3, Fs.az. ~tS4= A, o) 7 I,(continued) McGuire Units 1 and 2 5.6-2 Amendment Nos-T-M6-

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

3. Z Shutdown Bank Insertion Limit for Specification 3.1.5,

'. g. Control Bank Insertion Limits for Specification 3.1.6, Axial Flux Difference limits for Specification 3.2.3, &..X. Heat Flux Hot Channel Factor for Specification 3.2.1, Nuclear Enthalpy Rise Hot Channel Factor limits for Specification 3.2.2, Overtemperature and Overpower Delta T setpoint parameter values for Specification 3.3.1, Accumulator and Refueling Water Storage Tank boron concentration limits for Specification 3.5.1 and 3.5.4, / 11, ,Si Reactor Coolant System and refueling canal boron concentration limits for Specification 3.9.1, 17 1.. Spent fuel pool boron concentration limits for Specification 3.7.14, / 13$ X. SHUTDOWN MARGIN for Specification 3.1.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," Q Proprietary).

2.

WCAP-1 0266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE," (E Proprietary).

3.

BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B&W I.I (continued) McGuire Units 1 and 2 5.6-3 Amendment Nos. >344T t

ATTACHMENT 1B CATAWBA UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES MARKED COPY

INSERT 1 the COLR for four loop operation; and the following SLs shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 2 the 95/95 DNB criterion for the DNB correlation(s) specified in the thermal-hydraulic analytic method(s) listed in Section 5.6.5(b). 2.1.1.2 The peak fuel centerline temperature shall be maintained < the 95/95 CFM criterion for the fuel design(s) specified in the fuel mechanical analytic method(s) listed in Section 5.6.5(b). INSERT 2 The reactor core SLs are established to preclude violation of the following fuel design criteria: a.There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and

b. There must be at least a 95% probability at a 95%

confidence level that the hot fuel pellet in the core does not experience centerline fuel melting. The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Over Temperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and AI that the reactor

core SLs will be satisfied during steady state operation, normal operational transients, and AOOs. INSERT 3 The numerical limits of these variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow, based on previously analyzed maximum steam generator tube plugging, is retained in the TS LCO.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the specified in Fij1uf8A f 2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig. 2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour. 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes. Catawba Units I and 2 2.0-1 Amendment Nos.),1.E1

SLs 2.0 67bX DO NOT OPERATE IN THIS A 660 650 640 O 620 m /p a 610 600 / p\\ 590 ACCEPTABLE OPERATION 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits Four Loops in Operation Catawba Units 1 and 2 2.0-2 Amendment No.

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7) Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.3% (Unit 1) and 4.5% (Unit 2) of RTP. AT (1 + T S) (I1 ) *ATo{ K-K 2 ( + r S) [T (1I+ s) ]T' +K 3 (PP') f } Where: AT is the measured RCS AT by loop narrow range RTDs, 'F. ATO is the indicated AT at RTP, 'F. s is the Laplace transform operator, sec. T is the measured RCS average temperature, *F. J T is the nominal T,,, at RTP (allowed by Safety Analysis), < 6X401 P is the measured pressurizer pressure, psig Go d An d! coLE P is the nominal RCS operating pressure, = C&O-- K, = Overtemperature AT reactor NOMINAL TRIP SETPOINT, as presented in the COLR, K2 = Overtemperature AT reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR, x,, t2 = Time constants utilized in the lead-lag compensator for AT, as presented in the COLR, T3 = Time constant utilized in the lag compensator for AT, as presented in the

COLR,

¶4, T5 = Time constants utilized in the lead-lag compensator for Tan as presented in the COLR, = Time constant utilized in the measured Tang lag compensator, as presented in the COLR, and fi(Al) = a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup lests'such that: (i) for qt - qb between the "positive" and "negative" f1(Al) breakpoints as presented in the COLR; fl(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent Al that the magnitude of q, - qb is more negative than the f1(Al) "negative" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f1(AI) "negative" slope presented in the COLR; and (continued) Catawba Units 1 and 2 3.3.1 -18 Amendment Nos.J;9-

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7) Reactor Trip System Instrumentation (iii) for each percent Al that the magnitude of qt - qb is more positive than the f1(AI) "positive" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f1(AI) "positive" slope presented in the COLR. Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 2.6% (Unit 1) and 3.1% (Unit 2) of RTP. AT(I+TS (i l J<AT Ks-K rz.:s ( I T K4 TT-T f (AI)} Where: AT is the measured RCS AT by loop narrow range RTDs, "F. ATO is the indicated AT at RTP, 'F. s is the Laplace transform operator, sec1l. T is the measured RCS average temperature, "F. T is the nominal T., at RTP (calibratio pe o K4 = Overpower AT reactor NOMINAL TRIP SETPOINT as presented in the

COLR, K5

= A/tlfor increasn averaetrea for de asing average temperature, It F-Ur.=> Ace aen. K,6 = Overpower AT reactor trip heatup set cient as presented in the COLR for T > T and K6 =or T<T, T1, T2 = Time constants utilized in the lead-lag compensator for AT, as presented in the COLR, 13 = Time constant utilized in the lag compensator for AT, as presented in the

COLR,

= Time constant utilized in the measured T8,g lag compensator, as presented in the COLR, 17' = Time constant utilized in the rate-lag controller for Tayg, as presented in the COLR, and f2(AI) = a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for qt - qt between the "positive" and "negative" f2(Al) breakpoints as presented in the COLR; f2(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued) Catawba Units 1 and 2 3.3. 1-1 9 Amendment Nos. -

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits - -I LCO 3.4.1 APPLICABILITY: RCS DNB parameters for pressurizer pressuresRCS average temperature, 2

,tshalI be within the limits specified in rt_

G od e-5 1 ovFa me- -J MODE 1. .r =A.A a-NT-

---

--a----------- -- ---------- ---- --------

Pressurizer pressure limit does not apply during:

a.

THERMAL POWER ramp > 5% RTP per minute; or

b.

THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours RCS average to within limit. temperature DNB parameters not within limits. B. RCS total flow rate t 7Z- _ B.1 Reduce THERMAL POWER to

• 98% RTP.

AND 2 hours 6 hours B.2 Reduce the Power Range Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP. I, (continued) Catawba Units 1 and 2 3.4.1 -1 Amendment Nos. O f

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rate Vrfy-ri 0j Jo oF -Vc At W Syn pz - F,4 -'z / 0, _- C.1 Restore RCS total flow rate 2 hours I ORi LEw 1R X' C.2.1 e c T ERMAL POWER to < 50% RTP. AND C.2.2 Reduce the Power Range Neutron Flux - High Trip Setpoint to < 55% RTP. AND C.2.3 Restore RCS total flow rate 2 hours 6 hours 24 hours I

I 9 lc

- 0 v- -r,*W-0) --
(-1'0t-Q'.- D. Required Action and associated Completion Time not met. D.1 Be in NIODE 2. 6 hours Catawba Units 1 and 2 3.4.1 -2 Amendment Nos. 51. QL:t

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Veri ressurizer ressure is within limits 12 hours SR 3.4.1.2 Very avera ature is within limitsX 12 hours SR 3.4.1.3 Verify RCS total flow rate is within limitsK 12 hours SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total 18 months flow indicator. \\ '~A. ~' Catawba Units 1 and 2 3.4.1 -3 Amendment Nos..7W7¶~

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 fy Table 3.4.1-1 (page 1 of 1) RCS DNB Parameters

1.

Indicated RCS AM Temperature - Ur < 587.2 'F < 586.9 0F iputer < 587.7 'F < 587.5 'F Indicated RCS Average Temperature - Unit 2 4 3 meter < 592.9 'F < 592.6 0F 4 3 < 593.4 'F < 593.2 'F

2.

Indicated Pressurizer meter Pressure 4 3 > 2219.8 psig > 2222.1 psig I I )uter > 2215.8 psig > 2217.5 psig

3.

RCS Total Flow 2 390,000 gpm I Catawba Units 1 and 2 3.4.1 -4 Amendment Nos.18

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A0Os. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.

There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b.

The hot fuel pellet in the core must not experience centerline fuel melting. The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient f C conditions for Reactor Coolant System (RCS) temperaturepressure, and THE R level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities. Automatic enforcement of these reactor core SLs is provided b the < ~pf P/ A rj"U 0*to //- Th/litato a e eth9 in e trai> ur y e t 9 eg The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded. Catawba Units 1 and 2 B 2.1.1-2 Revision No.3

Reactor Core SLs B 2.1.1 BASES ) CL A SAFETY LIMITS The 4VA*V provided in VW eX,/fKishow the loci of points offA* RCS Pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature e AC 4 4 remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation. 2 APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER. SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 places VIOLATIONS the unit in a MODE in which this SL is not applicable. The allowed Completion Time of 1 hour recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage. REFERENCES

1.

10 CFR 50, Appendix A, GDC 10.

2.

UFSAR, Section 7.2. $ /- /a-E-po/1y'§ da¶1$9.6 .4 Catawba Units 1 and 2 B 2.1.1-3 Revision No.A25

Reactor Core SLs B 2.1.1 Ii 0. 0-. i SoM Safet T1vc (OF) OTAT CORE Figure B 2.1.1-1 Illustration of Overtemperature and Overpower AT Protection Catawba Units 1 and 2 B 2.1.1-4 Revision No.All"

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICAB Y (continued) setpoints are based on a minimum &~~f4loflow em '= g G o 5Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input. The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8. In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 10.b) because of the lower power level and the greater margin to the design limit DNBR.

b.

Reactor Coolant Flow-Low (Two Loops) The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow. Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The O > C aL setpoints are based on 'minimum P)Wviad flowgf$,' Each loop has three flow detectors to monitor flow. The flow signals are not used for any control system input. The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE. In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR. Catawba Units 1 and 2 B 3.3.1-18 Revision No. 3v

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) c The pressurizer pressure limits and the RCS average temperature limits correspond to analytical IimitsdZZWVV 4^u4 used in the safety analyses, with allowance for measurement uncertainty. The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2). LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses. -.- Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion. RCS total flow rate contains a measurement error 9YVD ' based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner. Therefore, a penalty40,V 1 for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance WIW,4for no fouling. The numerical values jvffi4{W/' for pressure and average /dt f s t temperatur re given for the measurement location with adjustments for C c 'L-P'-the indication instruments. APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern. A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations. Catawba Units 1 and 2 B 3.4.1-2 Revision No.>,

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued) is provided in SL 2.1.1, uri c' P 4D7oPIT s LDJ S "R eactor Core SLs." ,slVW1ware less restrictive than the limits of this Z 9 R LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded. ACTIONS A.1 Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s). The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience. '-4)/EW o itZ~ \\B.1 and B.2 RCS total flow rate is not a controllable parameter and is not expected to T x vary during steady state operation. If the indicated RCS total flow rate is 6jCj % O 0 4, iM

1. Y

then THERMAL POWER may not

/(At~ as_ rcF>. exceed 98% RTP. THERMAL POWER must be reduced within 2 hours. /J 7fl. 2 / The Completion Time of 2 hours is consistent with Required Action A.1. In addition, the Power Range Neutron Flux - High Trip Setpoint must be reduced from the nominal setpoint by 2% RTP within 6 hours. The Completion Time of 6 hours to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System. = C.1, C.2.1, C.2.2, and C.2.3 ,-J --i*E% Tog u'e %lIf the indicated RCS total flow rate is, then RCS total flow g y must be restored to ,~within 2 hours or power must be reduced to less than 50% RTP. The Completion Time of 2 hours is consistent with Required Action A.1. If THERMAL POWER is reduced to less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint Catawba Units 1 and 2 B 3.4.1-3 Revision No.X

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued) must also be reduced to < 55% RTP. The Completion Time of 6 hours to HARP T Ika) OR reset the trip setpoints is consistent with Required Action B.2. This is a L Eo p \\sensitive operation that may inadvertently trip the Reactor Protection TS1IC Ado.~' (r pSystem. Operation is permitted to continue provided the RCS total flow is LT estre t i within 24 hours. The Completion Time of 24 {a) bac-Go hours is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint. D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours. The Completion Time of 6 hours is reasonable to reach the required plant conditions in an orderly manner. SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. Catawba Units 1 and 2 B 3.4.1 -4 Revision Noxl

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) z.-. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3.1.3, c A

3 -XZ Shutdown Bank Insertion Uimit for Specification 3.1.5, A,

'i. 2 Control Bank Insertion Limits for Specification 3.1.6, X.. Axial Flux Difference limits for Specification 3.2.3, CA. £ Heat Flux Hot Channel Factor for Specification 3.2.1,

7. E.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3.2.2, X Overtemperature and Overpower Delta T setpoint parameter values for Specification 3.3.1, /$. ~ Accumulator and Refueling Water Storage Tank boron concentration limits for Specification 3.5.1 and 3.5.4, / (l 0. Reactor Coolant System and refueling canal boron concentration limits for Specification 3.9.1, TOZ. W. Spent fuel pool boron concentration limits for Specification 3.7.15, 13y, b SHUTDOWN MARGIN for Specification 3.1.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY' (at Proprietary).

2.

WCAP-1 0266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE" (W Proprietary). F-Y9O

3. t-4 (continued)

Catawba Units 1 and 2 5 6-3 Amendment Nos. 149

ATTACHMENT 2A McGUIRE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES (TO BE PROVIDED TO THE NRC UPON ISSUANCE OF APPROVED AMENDMENT)

ATTACHMENT 2B CATAWBA UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES (TO BE PROVIDED TO THE NRC UPON ISSUANCE OF APPROVED AMENDMENT)

ATTACHMENT 3 DESCRIPTION OF PROPOSED CHANGES AND TECHNICAL JUSTIFICATION

DESCRIPTION OF PROPOSED CHANGES AND TECHNICAL JUSTIFICATON Proposed Changes The proposed changes reduce the required minimum measured Reactor Coolant System (RCS) flow rate from 390,000 gpm to a previously reviewed and approved value of 382,000 gpm and relocate RCS related cycle-specific parameter limits from the Technical Specifications (TS) to, and thus expand, the Core Operating Limits Reports (COLR) for the McGuire and Catawba Nuclear Stations. McGuire and Catawba Units 1 and 2 were originally licensed with a minimum reactor coolant system (RCS) flow rate of 385,000 gpm. Due to steam generator tube plugging issues, analyses were performed for McGuire and Catawba Units 1 and 2 assuming a bounding tube plugging percentage and a license amendment to reduce the RCS minimum flow rate from 385,000 gpm to 382,000 gpm was requested for McGuire Units 1 and 2 and Catawba Unit 1. Although these analyses were applicable to Catawba Unit 2, a license amendment to lower the RCS minimum flow rate was not requested for this unit because Catawba Unit 2's steam generators had not experienced the same rate of tube plugging as had Catawba Unit 1. This license amendment request was approved by the NRC staff in Safety Evaluation Reports dated December 17, 1993 (for Catawba Unit 1)1 and March 22, 1994 (for McGuire Units 1 and 2)2. Subsequent to the replacement of steam generators, McGuire Unit 1 and 2 and Catawba Unit 1 were successfully operated with an RCS minimum total flow rate of 382,000 gpm in the technical specifications until those values were changed to 390,000 gpm through NRC Safety Evaluation Reports of March 2, 2000 for McGuire, and March 1, 2000 for Catawba. The NRC Safety Evaluation Report of March 1, 2000, also revised the Catawba Unit 2 RCS minimum total flow rate from 385,000 gpm to 390,000 gpm in the technical specifications. The analyses supporting the RCS minimum total flow rate of 390,000 gpm assumed a minimal steam generator tube plugging l Reference Duke submittal of October 25, 1993, as supplemented by letters of December 3 and 6, 1993 2 Reference Duke submittal of October 25, 1993, as supplemented by letters of December 3, 1993, and February 14, 1994

percentage. The RCS minimum total flow rates for McGuire and Catawba Units 1 and 2 were increased to make more effective use of available operating and analytical margins. This 390,000 gpm RCS total flow rate should be considered a cycle-specific minimum value, reflecting the condition of the McGuire and Catawba steam generators at the time the license amendment request was made. Values for the RCS pressure, temperature, and flow are specified in TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure for Nucleate Boiling (DNB) Limits." The proposed change reduces the required minimum measured RCS flow rate for McGuire and Catawba Units 1 and 2 from 390,000 gpm to 382,000 gpm, and replaces the current minimum RCS flow rate of 390,000 gpm with a reference to the COLR. The previously reviewed and approved accident evaluation of the replacement steam generators at a minimum RCS flow rate of 382,000 gpm was included as Attachment 1 to the McGuire Unit 1 and 23, and Catawba Unit 14 steam generator replacement license amendment requests of September 30, 1994. The proposed revision to TS 3.4.1, and associated Bases, further relocates Table 3.4.1-1, "RCS DNB Parameters," to the COLR. The 382,000 gpm RCS minimum flow rate, will be maintained in TS 3.4.1 so as to assure that a lower flow rate will not be used without prior NRC approval. TS 5.6.5, "Core Operating Limits Report (COLR)," has been modified to reflect the above relocations to the COLR. The proposed amendment also relocates TS Figure 2.1.1-1, "Reactor Core Safety Limits," to the COLRs, replacing it with fuel integrity limits (95/95 DNB and 95/95 CFM 3 Reference Duke submittal of September 30, 1994, as supplemented by letters of September 18, 1995, and March 15, April 29, May 16, September 23, and October 28, 1996, and January 16, April 22, and May 2, 1997, and the NRC Safety Evaluation Report dated May 5, 1997, for McGuire Units 1 and 2 4 Reference Duke submittal of September 30, 1994, as supplemented by letters of September 18, 1995, January 19, March 15, May 16, and August 27, 1996; and the NRC Safety Evaluation Report of August 29, 1996, for Catawba Unit 1

criteria) to reflect McGuire and Catawba's use of different fuel designs from various vendors and to better represent the safety limits requirement to 10CFR §50.36. As discussed in the Safety Evaluation Reports to WCAP-14483-A, it is necessary to relocate TS Figure 2.1.1-1 to the COLR since cycle-dependent changes to parameters upon which TS Figure 2.1.1-1 is based would require a license amendment request to revise the figure. The amendment also revises TS Table 3.3.1-1, "Reactor Trip System Instrumentation," by relocating numerical values pertaining to Overtemperature AT and Overpower AT nominal RCS operating pressure, nominal Tavg, and constant (K) values to the COLR. Additionally, the basis for the RCS low flow reactor trip setpoints has been revised by referring to the COLR in the associated Bases document. TS 5.6.5, "Core Operating Limits Report (COLR)," will be modified to reflect the above relocations to the COLR. The proposed changes will allow Duke the flexibility of enhancing operating and core design margins without the need for cycle-specific license amendment requests. The relocation of these cycle-specific TS values to the COLR will result in a more complete COLR containing cycle-specific operating conditions and core reload related parameters. The safety and quality of operations at Duke's McGuire and Catawba nuclear stations will not be compromised by the implementation of this amendment request as TS 5.6.5(c) requires that all applicable limits of the safety analyses be met when generating cycle-specific requirements in the COLR. Basis for Proposed Change NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameters From Technical Specifications," dated October 4, 1988, provides guidance to licensees for the removal of cycle-dependent variables from the TS provided that these values are included in a COLR and are determined with NRC-approved methodologies referenced in the TS. Westinghouse Electric Company (Westinghouse) subsequently developed WCAP-14483, "Generic Methodology for Expanding Core Operating Limits Report," describing how cycle-specific

parameters may be relocated to the COLR. WCAP-14483 was accepted for referencing by the NRC on January 19, 1999. The Safety Evaluation Report, contained in the January 19, 1999 NRC letter approving WCAP-14483-A, concludes that additional information contained in the TS may be relocated to the COLR. The limits on the parameters which are removed from the Technical Specifications and added to the COLRs must be developed or justified using NRC-approved methodologies. All accident analyses, performed in accordance with these methodologies, must meet the applicable NRC-approved limits of the safety analysis. The removal of parameter limits from the Technical Specifications and their addition to the COLRs does not obviate the requirement to operate within these limits. Furthermore, any changes to those limits must be performed in accordance with TS 5.6.5(c). If any of the applicable limits of the safety analyses are not met, prior NRC approval of the change is required, just as is the case for a license amendment request. For more routine modifications, where NRC-approved methodologies and limits of the safety analysis remain applicable, the potentially burdensome and lengthy process of amending the Technical Specifications may be avoided. The requested changes are essentially administrative in nature; therefore, the required level of safety will be maintained. The requested changes are based upon NRC approved Westinghouse Owners Group (WOG) Technical Specifications Task Force (TSTF) TSTF-339, "Relocate TS Parameters to the COLR Consistent with WCAP-14483,0 Revision 2, and Westinghouse WCAP-14483-A. In accordance with these documents, previously approved RCS minimum total flow rates are retained in the TS to preclude the use of lower flow rates without prior NRC approval. The return to a lower RCS minimum flow rate, previously reviewed and approved by the NRC, is similar to that submitted by the Comanche Peak Nuclear Station on May 24, 1999, and approved by the NRC in the Safety Evaluation Report dated August 30, 1999. Consistent with the Comanche Peak submittal is a generic statement in TS 2.1.1 requiring compliance with the DNBR limit for the correlation(s) used for a specific core design. The same approach has also been made with respect to the CFM limits. Both of these statements more clearly address the requirements of 10CFR §50.36 by stating the

actual safety limits of DNB and CFM. Future changes to TS 2.1.1 are also minimized.

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION As required by 10CFR50.91(a)(1), this analysis is provided to demonstrate that the proposed license amendment does not involve a significant hazard. Conformance of the proposed amendment to the standards for a determination of no significant hazards, as defined in 10CFR50.92, is shown in the following:

1) Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The relocation of Reactor Coolant System (RCS) related cycle-specific parameter limits from the Technical Specifications (TS) to the Core Operating Limits Reports (COLR) proposed by this amendment request does not result in the alteration of the design, material, or construction standards that were applicable prior to the change. The proposed change will not result in the modification of any system interface that would increase the likelihood of an accident since these events are independent of the proposed change. The proposed amendment will not change, degrade, or prevent actions, or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the UFSARs. Therefore, the proposed amendment does not result in the increase in the probability or consequences of an accident previously evaluated.

2) Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. This change does not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the facility which should introduce any new accident causal mechanisms. This amendment request does not impact any plant systems that are accident initiators.

3) Does the proposed change involve a significant reduction in margin of safety?

No. Implementation of this amendment would not involve a significant reduction in the margin of safety. Previously approved methodologies will continue to be used in the determination of cycle-specific core operating limits appearing in the COLRs. Additionally, previously approved RCS minimum total flow rates for McGuire and Catawba are retained in their respective TS so as to assure that lower flow rates will not be used without prior NRC approval. Consequently, no safety margins will be impacted. Based on the above, it is concluded that the proposed license amendment request does not result in a reduction in margin with respect to plant safety. Conclusion Based on the preceding analysis, it is concluded that the relocation of Reactor Coolant System (RCS) related cycle-specific parameter limits from the TS to the COLR does not involve a Significant Hazards Consideration Finding as defined in 10CFR50.92.

ATTACHMENT 5 ENVIRONMENTAL ANALYSIS

ENVIRONMENTAL ANALYSIS The proposed amendment has been reviewed against the criteria of 10CFR51.22 for environmental considerations. The proposed amendment does not involve a significant hazards consideration, increase the types and amounts of effluents that may be released offsite, or result in the increase of individual or cumulative occupational radiation exposures. Therefore, the proposed amendment meets the criteria provided by IOCFR51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.}}