View source for CNS-17-030, Catawba, Unit 1, Submittal of Core Operating Limits Report (Colr) for Cycle 24 Reload Core

{{Adams
| number = ML17179A428
| issue date = 06/20/2017
| title = Catawba, Unit 1, Submittal of Core Operating Limits Report (Colr) for Cycle 24 Reload Core
| author name = Simril T
| author affiliation = Duke Energy Carolinas, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000413
| license number = NPF-035
| contact person = 
| case reference number = CNC-1553.05-00-0651 , Rev. 0, CNS-17-030
| document report number = CNEI-0400-310, Rev 0
| document type = Fuel Cycle Reload Report, Letter type:CNS
| page count = 34
}}

=Text=
{{#Wiki_filter:CNS-17-030 June 20 , 2017 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington , DC 20555-0001 

==Subject:==
Duke Energy Carolinas , LLC (Duke Energy) Catawba Nuclear Station , Unit 1 Facility Operat i ng License Number NPF-35 Docket Number 50-413 Tom Si mri l Vice P resident Catawba Nuclear Station Du ke En er gy C N01VP I 4800 Concord Road York , SC 29745 o: 803.701.3340 f: 803. 7 0 1.3221 tom.simr i l@duke-energy
.com Core Operating Limits Report (COLR) for Cycle 24 Reload Core Pursuant to Catawba Technical Specification 5.6.5d., please find attached an information copy and an electronic copy of the subject COLR. This COLR revision is being submitted to update the limits of the Unit 1 Cycle 24 reload core. The electronic copy of this COLR is included with this letter in portable document format (PDF) on a digital versatile disc (DVD). The electronic copy includes the power distribution monitoring factors. This letter , the enclosed COLR , and the included DVD COLR with Attachment A , do not contain any regulatory commitments. Please direct any questions or concerns to Carrie Wilson , Sr. Engineer , at (803) 701-3014. Sincerely , Tom Simril Vice President , Catawba Nuclear Station Enclosures (paper COLR and DVD COLR with Attachment A) www.duke-energy.com U.S. Nuclear Regulatory Commission CNS-17-030 Page 12 J une 20 , 2017 xc (with enclosures)
: C. Haney, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta , GA 30303-1257 J. D. Austin , Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station M. Mahoney , Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop 0-8H4A R o ckville , MD 20852 Enclosures Catawba Unit 1 Cycle 24 COLR (paper COLR and DVD COLR with Attachment A)
Catawba I Cycle 24 Core Operating Limits Report Revision 0 Apri l 2017 

==Reference:==

CNC-1553.05-00-0651 , Rev. 0 Duke Energy Carolinas, LLC QA CONDITION I CNEI-0400-310 Page I Revision 0 The information presented in this report ha s been prepared and issued in accordance with Catawba Technical Specification 5.6.5.
Catawba 1 Cycle 24 Core Operating Limits Report Implementation Instructions for Revision 0 Revision Description and CR Tracking CNEI-0400-310 Page 2 Revision 0 Revision 0 of the Catawba Unit 1 Cycle 24 Core Operating Limits Report (COLR) contains limits specific to the reload core. There is no CR associated with this revision.
Implementation Schedule The Catawba Unit 1 Cycle 24 COLR requires the reload 50.59 be approved prior to implementation and fuel loading. Revision 0 may become effective any time during No MODE between cycles 23 and 24 but must become effective prior to entering MODE 6 which starts cycle 24. The Catawba Unit 1 Cycle 24 COLR will cease to be effective during No MODE between cycle 24 and 25. Data files to be Implemented No data files are transmitted as part of this document.
Revision 0 CNEI-0400-310 Page 3 Revision 0 Catawba 1 Cycle 24 Core Operating Limits Report Effective Date April 2017 REVlSION LOG Pages Affected 1-31 , Appendix A* COLR ClC24 COLR , Rev. 0 *Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance. Appendix A is included only in the electronic COLR copy sent to the NRC.
Catawba 1 Cycle 24 Core Operating Limits Report CNEI-0400-3 1 0 Pa ge 4 Rev i s i on 0 1.0 Core Operating Limits Report TS Sect ion 2.1.1 3.1.1 3.1.3 3.1.4 3.1.5 3.1.6 3.1.8 3.2.1 3.2.2 3.2.3 3.3.1 3.3.9 3.4.1 3.5.1 3.5.4 3.7.15 3.9.1 5.6.5 SLC Section 1 6.7-9 16.9-11 16.9-1 2 This Core Operat in g Lim it s Report (COLR) h as been prepared in accorda n ce wit h requirement s of Technical Spec i fication 5.6.5. Tec hni cal Spec ifi cations that referenc e this report are l isted below a l ong wit h the NRC approved analytica l methods used to develop and/or determin e COLR parameters identified in Technical Specifications.
COL R NRC A pproved Tec hnical S p ec ification s COL R Parameter Section Methodology (Section 1.1 Number) Reactor Core Safety L imit s RCS Temperature and Pres sure 2.1 I 6 , 7, 8 , 9 , 10 , 1 2 , 15 , Safety Limits 1 6 , 19 , 20 Margin S_!!utdo wn Marg i n 2.2 6 , 7 , 1 2 , 14! 1 5 , 1 6 , 19 , 20 -
Tell]perature Coefficie!1t . MTC 2.3 8 , 1 4 , 1 6 , 1 8 Rod Group A li g nm ent Limit s S hut down Margin 2.2 6 , 7 , 8 , 1 2 , 14 , 1 5 , 1 6 , 19 , 20 Shutdown Bank In s ertio n Limit Shutdown Margin 2.2 2 , 4 , 6 , 7 , 8 , 9 , 1 0 Rod In sert ion Limits 2.4 1 2 , .!_4 , l5 , 1 6 , 1 9 , 20 Control Bank In s ertion Limit Shutdown Margin 2.2 2 , 4 , 6 , 7 , 8 , 9 , 10 Rod In sert ion Limits 2.5 1 6 , 1 9 , 20 --
Exceptions S hutdo wn Margin 2.2 6 , 7 , 8 , Jl , 1 4 , 1 5 , 1 6 , 1 9 , 20 *--H eat F l ux Hot Channe l Factor Fo I 2.6 2 , 4 , 6 , 7 , 8 , 9 , 1 0 , AFD 2.8 1 2, 15 , 16 , 1 9 , 20 OT.t-.T 2.9 Penalty Factors I 2.6 Nuclear En th a l py Rise Hot Channel FiiH 2.7 2 , 4 , 6 , 7 , 8 , 9 , 1 0 Factor Pena l ty Facto r s 2.7 12_,_!5!..16! 1 9 1 20 Axial Flux Difference AFD 2.8 2_, 4 2..6 ,_7 , 8 , 1 6 --Reactor Trip System In s trum e ntation OT.t-.T 2.9 6 , 7 , 8 , 9 , 10 , 1 2 OP.t.T 2.9 15 , 16 , 19 , 20 Boron Dilution Mitigation System R eactor Makeup Water F l ow R ate 2.10 6 , 7 , 8 , 14 , 1 6 RCS Pr ess ure , Temperature and F l ow R CS Pr ess ur e , Temperatu r e and 2.11 6 , 7 , 8 , 9 , 1 0 , 1 2 , limit s for DNB F l ow ) ---' Accumulators Max and M in Boron Cone. 2.1 2 6, 7 !!., 14 , 1 6 Refu s: Iin g Water Storage Tank Max and M in Boron Co n e. 1 2.1 3 6, 7,_8 , 14 , 1 6 Fuel Pool Boron M in Boron Concentration J 2.1 4 2, 7, 8 , 1 4 , 1 6 Refu e li ng Operat i ons -Boron M in Boron Concentratio n 2.15 6 , 7 , 8 , 1 4 , 1 6 Concentration Core Operating Limits R eport (COL R) A n a l ytica l Methods I. I None The Selected Licensee Commitments that reference this r eport are listed below COL R NRC A pproved Se lected Licensee Co mmitment COLR Parameter Sect ion Methodology (Section I.I Number) Sta ndb y Shutdown System Standby Mak e up Pump Wate_r S uppl y 2.16 I 6, 7 , 8 , 1 4 , 1 6 Boration Syste m s -Borated Water Bo rate d Water Vo lum e an d Co n e. for 2.17 I 6 , 7 , 8 , 1 4 , 1 6 Source -Shutdown B AT/RWST Boration Systems -Borated Water Borat ed Water Volume and Co n e. for 2.18 6 , 7 , 8 , 1 4 , 1 6 So ur ce -Operating BAT/RWST Catawba 1 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods CNEl-0400-310 Pa ge 5 Revision 0 Analytical methods used to determine core operating limits for parameter s identified in Technical Specifications and previously reviewed and approved by the NRC as specified in Technical Specification 5.6.5 are as follows. 1. WCAP-9272-P-A , " Westinghouse R e load Safety Evaluation Methodology
," (W Proprietar y). R ev ision 0 Report Date: July 1985 Not Used 2. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code, " (W Proprietary). Re vis ion 0 Report Date: August 1985 Addendum 2 , " Addendum to the We st inghouse Small Break ECCS Evaluation Model U s in g the NOTRUMP Code: Safety Injection into the Broken Loop and COS! Condensation Model ," (W Proprietary). (Referenced in Duke Letter DPC-06-I 01) Revision I Report Date: July I 997 3. WCAP-I 0266-P-A , "T he I 981 Version of Westinghouse Evaluation Model Using BASH Code", (W Propri etary). Revi s ion 2 Report Date: March 1987 Not Used 4. WCAP-12945-P-A , Volume I and Volumes 2-5 , " Code Qualification Document for Estimate Los s of Coolant Analysis ," (W Proprietary). R evisio n: Volume 1 (Revision 
: 2) and Volumes 2-5 (Revision I) Report Date: March 1998 5. BAW-10168P-A , " B&W Loss-of-Coolant Accident Evaluation Model for Recirculatin g Steam Generator Plants ," (B&W Proprietary). Revision I SER Date: January 22 , 1991 Revision 2 SER Dates: August 22, 1996 and November 26 , 1996 Revision 3 SER Date: June I 5 , 1994 Not Used Catawba 1 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods (continued)
CNEI-0400-310 Page 6 Revision 0 6. DPC-NE-3000-PA , " Thermal-Hydraulic Transient Analysis Methodology
," (DPC Proprietary).
Revision Sa Report Date: October 2012 7. DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology
," (DPC Proprietary).
Revision I Report Date: March 2015 8. DPC-NE-3002-A , "UFSAR Chapter 15 System Transient Analysis Methodology".
Revision 4b Report Date: September 201 0 9. DPC-NE-2004P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01 ," (DPC Proprietary).
Revision 2a Report Date: December 2008 10. DPC-NE-2005P-A , "Thermal Hydraulic Statistical Core Design Methodology
," (DPC Proprietary). Revision 5 Report Date: March 2016 11. DPC-NE-2008P-A , "Fuel Mechanical Reload Analysis Methodology Using T AC03 ," (DPC Proprietary). Revision 0 Report Date: April 3 , 1995 Not Used 12. DPC-NE-2009-PA , "Westinghouse Fuel Transition Report," (DPC Proprietary).
Revision 3c Report Date: March 2017 13. DPC-NE-1004-A , "Nuclear Design Methodology Using CASM0-3/SfMULATE-3P." Revision la Report Date: January 2009 Not Used Catawba 1 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods (continued)
CNEI-0400-310 P age 7 Revi s ion 0 14. DPC-NF-2010-A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design." Revision 2a Report Date: December 2009 15. DPC-NE-2011-PA , "Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors ," (DPC Proprietary).
Revision l a Report Date: June 2009 16. DPC-NE-1005-PA , "Nuclear Design Methodology Using CASM0-4 I SIMULATE-3 MOX" , (DPC Proprietary). Revision 1 Report Date: Nove mb er 12 , 2008 17. BA W-10231 P-A , "COP ERN IC Fuel Rod De sign Computer Code" (Framatome ANP Proprietary)
Revision I SER Date: January 14 , 2004 Not Used 18. DPC-NE-1007-PA , " Conditional Exemptio n ofthe EOC MTC Measurement Methodology
'', (DPC and W Proprietary)
Revision 0 Report Date: A pri I 2015 19. WCAP-12610-P-A , " VANTAGE+ Fuel Assembly Reference Core Report ," (W Proprietary).
Revision 0 Report D ate: Apr il 1995 20. WCAP-12610-P-A 
& CENPD-404-P-A , Addendum 1-A , " Optimized ZIRLO""," (W Propri eta r y). R evisio n 0 Report D ate: July 2006 Catawba 1 Cycle 24 Core Operating Limits Report 2.0 Operating Limits CNEI-0400-310 Page 8 Revision 0 Cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using NRC approved methodologies specified in Section 1.1. 2.1 Reactor Core Safety Limits (TS 2.1.1) The Reactor Core Safety Limits are shown in Figure 1. 2.2 Shutdown Margin -SDM (TS 3.1.1, TS 3.1.4, TS 3.1.5, TS 3.1.6, TS 3.1.8) 2.2.1 For TS 3.1. l , SOM shall be greater than or equal to 1.3%
in MODE 2 with Keff < 1.0 and in MODES 3 and 4. 2.2.2 For TS 3.1.1 , SOM shall be greater than or equal to 1.0% .6K/K in MODE 5. 2.2.3 For TS 3.1.4, SOM shall be greater than or equal to 1.3% .6K/K in MODE 1 and MODE2. 2.2.4 For TS 3.1.5 , SOM shall be greater than or equal to 1.3% .6K/K in MODE 1 and MODE 2 with any control bank not fully inserted. 2.2.5 For TS 3.1.6 , SOM shall be greater than or equal to 1.3% .6K/K in MODE 1 and MODE 2 with 1.0. 2.2.6 For TS 3.1.8 , SOM shall be greater than or equal to 1.3% .6KIK in MODE 2 during PHYSICS TESTS.
660 500 580 0.0 Catawba 1 Cycle 24 Core Operating Limits Report Figure 1 R e actor Core Safety Limit s Four Loop s in Operation CNE I-0 400-3 1 0 Page 9 R ev i sio n 0 DO NO OP RATE I N T H IS AA. A ACC_PTAB lE OPERATIC 02 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Powe Catawba 1 Cycle 24 Core Operating Limits Report 2.3 Moderator Temperature Coefficient 
-MTC (TS 3.1.3) 2.3.1 Moderator Temperature Coefficient (MTC) Limit s are: CNEI-0400-310 Page 10 Revi s ion 0 MTC shall be less positive than the upper limits shown in Fig ure 2. BOC , ARO , HZP MTC s hall be less positive than 0.7 E-04 LiK/K/&deg;F. E OC , ARO , RTP MTC shall be les s ne gative than the -4.3 E-04 LiKIK/&deg;F lower MTC limit. 2.3.2 300 ppm MTC Surveillance Limit is: Measured 300 PPM ARO , equilibrium RTP MTC s hall be le ss negative than or equal to -3.65E-04 LiK/K/&deg;F. 2.3.3 The Revis e d Predicted near-E OC 300 ppm ARO RTP MTC s hall be calculated usin g th e procedure contained in DPC-NE-1007-PA. If the Re v ised Predicted MTC is less negative than or equal to the 300 ppm SR 3. l .3.2 Surveillance Limit , and all benchmark data contained in the surveillance procedure is satisfied , then a MTC mea s urement in accordance with SR 3.1.3.2 i s not required to b e performed. 2.3.4 60 PPM MTC Surveillance Limit is: Measured 60 PPM ARO , equilibrium RTP MTC shall be les s negative than or equal to -4.125E-04 LiK/K/&deg;F. Where: BO C = Beginnin g of Cycle (burn up corresponding to mo st po s itive MTC) EOC = E nd of Cycle ARO= All Rod s Out H ZP = Hot Zero Thermal Power RTP = Rated Thermal Po we r PPM= Parts per million (Boron) 2.4 Shutdown Bank Insertion Limits (TS 3.1.5) 2.4.1 Each shutdown bank s hall be w ithdrawn to at least 222 steps. Shutdown banks a re withdrawn in sequence and with no overlap. 2.5 Control Bank Insertion Limits (TS 3.1.6) 2.5.1 Control banks shall be within the insertion , sequence , and overlap limits shown in Figure 3. Specific control bank withdrawal and overlap limits as a function of the fully withdrawn po sitio n are shown in Table 1.
Catawba 1 Cycle 24 Core Operating Limits Report Figure 2 CNEI-0400-310 Pag e 11 Revi sio n 0 Moderator Temperature Coefficient Upper Limit Versus Power Leve l 1.0 Unacce ptabl e Operation Acce pt a bl e Operation 0.1 0.0 0 10 20 3 0 40 50 60 70 80 90 100 Percent of Rated Thermal Power NOTE: Co mp l iance w ith Technical Specification 3.1.3 ma y require rod withdrawal limits. Refer to th e U nit 1 ROD manual for detail s.
231 220 200 ,-., 180 ... "O .c 160 ... :i Catawba 1 Cycle 24 Core Operating Limits Report Figure 3 CNEI-0400-310 Pa ge 12 Revi s ion 0 Control Bank Insertion Limits Versus Percent Rated Therma l Power F ull y Withdrawn (Maximum=23 ------------------/ ,,. / / / / / / / Fully Withdrawn 
/ / Co ntr o l Bank 8 (Minimum=222) / / / / : c100%, 1 6 1) F / / / B (0%, 1 63) I / / / / / / 140 / / / 1 Co ntrolB a nk C 1 / ... rfJ ....... c: 0 E "' 0 Q., c: 0 '.C ... "' c: -"O 0 / / I I / 120 / / v / / / / / 100 / ,,. / / / / 80 v / / Co ntrol Bank D v / / / 60 / / ,,. / / 40 (0%, 47) / / / 20 -]F ull y In se rt e d 1 / / I 1 (30%, 0) I / 0 -I I/ 0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power The Rod Insertion Limits (RIL) for Control Bank D (CD), Contro l Bank C (CC), and Contro l Bank B (CB) can be ca l cu l ated b y: Bank C D RIL = 2.3 (P)-69 {30 <.::.P <.::.JOO}
Bank CC RIL = 2.3 (P) +47 {O <.::.P <.::. 7 6.J} for CC RIL = 222 {7 6.J < P <.::.JOO}
Bank CB RIL = 2.3(P) + J63 {O <.::. P <.::. 2 5. 7) for C B RIL = 222 {25. 7 < P <.::.JOO}
w her e P = %Rat ed Thermal Po we r NOTES: Co mpliance with Techn ical Specification 3.1.3 may require rod withdrawal limit s. Refer to the Un it 1 ROD manual for details.
Catawba 1 Cycle 24 Core Operating Limits Report Table 1 Control Bank Withdrawal Steps and Sequence F ull y Withdrawn a t 222 S tep s Co ntr o l Bank A 0 S t art 11 6 222 Stop 222 222 222 222 Co ntrol Bank B 0 0 S t a rt 1 06 11 6 222 Stop 222 222 Co ntrol Co ntr o l Bank C Bank D 0 0 0 0 0 0 0 Start 0 106 0 11 6 0 S t art 222 S t o p 1 06 Full y Withdrawn at 224 S t e p s C ontrol Bank A 0 Start 11 6 22 4 S t op 224 224 224 224 Co ntrol Bank B 0 0 S t a rt 1 08 11 6 224 S t o p 224 224 Co ntr o l C ontrol B a nk C B a nk D 0 0 0 0 0 0 0 S t a rt 0 1 08 0 11 6 0 Start 224 Sto p 1 08 Fully Withdrawn at 226 S t e ps Co ntr o l Bank A O Start 11 6 226 S t op 226 226 226 226 C ontrol Bank B 0 0 S t art 1 1 0 11 6 226 Stop 226 226 Co ntr o l C ontrol B a nk C Bank D 0 0 0 0 0 0 0 Start O 11 0 0 11 6 0 Start 226 Sto p 11 0 Full y Withdrawn a t 228 S tep s Control C ontrol Control Co ntrol Bank A Bank B B a nk C Bank D 0 Start 11 6 228 S t op 228 228 228 228 0 0 Start 11 2 11 6 228 S t o p 228 228 0 0 0 0 Start 11 2 11 6 228 S t o p 0 0 0 0 0 0 S t a rt 11 2 Full y Withdrawn at 230 S tep s Co ntrol Bank A 0 Start 11 6 230 Stop 2 30 230 230 230 Co ntrol Bank B 0 0 Sta rt 1 1 4 11 6 230 S t op 230 230 Control Co ntrol Bank C B a nk D 0 0 0 0 0 0 0 Start 0 11 4 0 11 6 0 S t a rt 230 Stop 11 4 Fully Withdrawn at 223 S tep s C ontrol BankA 0 S t a rt 11 6 223 Stop 223 223 223 223 Con t ro l Bank B 0 0 Start 1 07 11 6 223 Stop 223 223 C ontrol Bank C 0 0 0 0 Start 1 07 11 6 223 S t op C ontrol Bank D 0 0 0 0 0 0 S t a rt 1 07 Fully Withdr aw n at 22 5 Steps Co ntrol B a nk A 0 Sta rt 11 6 225 Stop 225 225 225 225 Co ntrol Bank B 0 O S t art 1 09 11 6 225 Stop 225 225 C ontrol Bank C 0 0 0 0 Start 1 09 11 6 225 Sto p Co ntrol Bank D 0 0 0 0 0 0 S t art 1 09 Fully Withdrawn at 22 7 Steps Co ntrol B a nk A 0 Start 11 6 227 Sto p 227 227 227 227 Co ntr o l Bank B 0 0 S t art I l l 11 6 227 Stop 227 227 C ontrol Bank C 0 0 0 0 Start Ill 11 6 227 Stop Co ntrol Bank D 0 0 0 0 0 0 Start II I Fully Withdrawn at 229 S tep s Co ntrol Bank A 0 S t a rt 11 6 229 Stop 229 229 229 229 C ontrol Bank B 0 0 Start 113 11 6 229 Stop 229 229 Co ntrol BankC 0 0 0 0 S t art 11 3 11 6 229 S t op C ontrol Bank D 0 0 0 0 0 0 S t a rt 11 3 Full y Withdrawn at 231 S teps Co ntrol Bank A 0 Start 11 6 23 1 Stop 23 1 23 1 23 1 23 1 Co ntrol Bank B 0 0 Start 11 5 11 6 23 1 S t op 23 1 23 1 Co ntrol BankC 0 0 0 O S t art 11 5 11 6 23 1 S t o p C ontrol Bank D 0 0 0 0 0 0 S t a rt 115 CNEI-0400-310 Page 13 Revi s ion 0 Catawba 1 Cycle 24 Core Operating Limits Report CNEI-0400-310 Page 14 Revision 0 2.6 Heat Flux Hot Channel Factor -FQ(X,Y,Z) (TS 3.2.1) 2.6.1 F Q (X , Y , Z) steady-state limit s are defined by the following relationships:
where , F R TP *K(Z)/P Q F *K(Z)/0.5 P = (Thermal Power)/(Rated Power) for P > 0.5 for P 0.5 Note: The measured FQ(X , Y , Z) shall be increased by 3% to account for manufacturing tolerances and 5% to account for measurement uncertainty w h en comparing against the LCO lim it. The manufacturing to leranc e and measurement un certai nt y are implicitly includ ed in the F Q s urv e ill ance limits as defined for COLR Sections 2.6.5 and 2.6.6. RTP 2.6.2 F Q = 2.70 x K(BU) 2.6.3 K(Z) is the norm al i zed F Q (X , Y , Z) as a function of core hei g ht. K(Z) for Westinghouse RFA fuel i s provided in Figure 4. 2.6.4 K(BU) is the norm alized F Q (X , Y , Z) as a function of burnup.
w ith the K(BU) penalty for Westinghouse RF A fuel is ana l ytica ll y confirmed in specific reload calculations. K(BU) i s set to 1.0 at all burnups. T h e following parameters are required for core monitoring per the Surve ill ance Requirements of Tech nic a l Spec ific ation 3.2. l: 2.6.5 L Fg(X , Y , Z)
* M Q (X , Y , Z) [FQ(X , Y , Z)]O P = UMT *MT* TILT where: (X , Y , Z)]O P = Cyc l e dependent maximum allowable design peaking factor F t (X , Y , Z) that e n sures F Q (X , Y , Z) LOCA limit i s not exceeded for operat i on within the AFD , RIL , and QPTR limits. (X , Y , Z)0 r includes a ll owances for calculation and measurement uncertainties.
Design po wer distribution for FQ. F t (X , Y , Z) is provided in Append i x Table A-1 for normal operati n g conditions and in Append i x Table A-4 for power escalation testing during initial startu p operation.
2.6.6 CNEI-0400-310 Pa ge 15 Revision 0 Catawba 1 Cycle 24 Core Operating Limits Report M Q (X , Y , Z) = Margin remaining in core location X,Y , Z to the LOCA limit in the transient power distribution.
M Q(X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in Appendix Table A-4 for pow er esca lation testing durin g initial star tup operation.
UMT Total Peak Measurement Uncertainty. (UMT = 1.05) MT E n g ineerin g Hot Channel Factor. (MT= 1.03). TILT Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.0 2. (TILT= 1.035) L RP S Fg(X,Y,Z)
* M c (X , Y , Z) [F Q (X , Y , Z)] = UMT *MT* TILT w h ere: = Cycle dependent maximum allowable de sig n peakin g factor that ensures FQ(X , Y ,Z) Centerline Fuel Melt (CFM) limit is not excee ded for operation within the AFD , RIL , and QPTR limits.
include s allowances for calculation and measurement uncertaintie
: s. D FQ(X,Y,Z)
Defined in Section 2.6.5. M c (X , Y , Z) Margin remaining to the CFM limit in core location X , Y , Z from the tran s ient po wer distribution. Mc(X , Y,Z) is provided in Appendix Table A-2 for normal operating conditions and in Appendix Table A-5 for power escalation testing during initial sta rtup operations.
UMT Defined in Section 2.6.5. MT = D efine d in Section 2.6.5. TILT = Defined in Section 2.6.5.
Catawba 1 Cycle 24 Core Operating Limits Report 2.6.7 KSLOPE = 0.0725 where: CNEI-0400-310 Pa g e 16 Revision 0 KSLOPE =Adjustment to Ki value from trip setpoint required to RP S compensate for each 1 % F;1 (X , Y , Z) exceeds (X , Y , Z) . 2.6.8 F Q (X , Y , Z) Penalty Factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.
Catawba 1 Cycle 24 Core Operat i ng Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westinghouse RFA Fuel CNE I-0400-3 10 P age 1 7 R e vi s i o n 0 1.2 00 (0.0 , 1.00) ( 4.0 , 1.00) 1.000 ..,__ ______ .... , 0.800 @: 0.600 0.400 0.2 00 0.000 0.0 C ore Hei g ht 0.0 ::;: 4 >4 1 2.0 2.0 ( 4.0 , 0.9259) K{Z) 1.000 1.000 0.9259 0.9259 4.0 6.0 Co r e H eig ht (ft) (1 2.0 , 0.9259) 8.0 1 0.0 1 2.0 Catawba 1 Cycle 24 Core Operating Limits Report Table 2 FQ(X,Y,Z) and F 6 tt(X,Y) Penalty Factors CNEI-0400-3 I 0 Pag e 18 Revision 0 For Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burn up (EFPD) 4 12 25 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 490 506 516 521 531 541 FQ(X,Y,Z)
Penalty Factor(%)
2.00 2.01 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 F , rn(X,Y) Penalty Factor (%) 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 Note: Linear int er polation is adequate for intermediate cycle burnup s. Al l cycle burnups outside the range of the table shall use a 2% pena l ty factor for both FQ(X , Y , Z) and F L\H(X , Y) for compliance with the Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.
Catawba 1 Cycle 24 Core Operating Limits Report CNEI-0400-310 Page 19 Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor -F AfI(X,Y) (TS 3.2.2) F steady-state limit s referred to in Technical Specification 3.2.2 are defined by the fo ll owing relationship. where: [F;H (X , Y)]L c o is the steady-state , maximum allowed radial peak and includes allowance s for calculation
/measurement uncertainty.
MARP(X , Y) = Cyc l e-specific operating limit Maxim um Allowable Radial Peaks. MARP(X , Y) radial peaking limi ts are provided in Tab l e 3. p Thermal Power Rated Thermal Power RRH =Therma l Power reduction required to compensate for each 1 % the m e asured radial peak , F:i (X , Y), exceeds the limit. (RRH = 3.34 , 0.0 < P :S 1.0) The following parameters are required for core monitoring per the survei ll ance requirements of Techn i cal Specification 3.2.2. 2.7.2 L SU RV (X , Y)] (X , Y) * (X , Y) UMR *TILT where: S URV (X , Y)] = Cycle dependent maximum allowa bl e design peaking factor that ensures the F tt<X , Y) limit is not exceeded for operation SU RV withi n the AFD , RIL , and QPTR limits. (X , Y) includes allowa nc es for calculation and measurement uncertainty. (X , Y) = Design radial power distribution for F (X , Y) is provided in Appendix Table A-3 for norm a l operatio n and in Append i x Tab l e A-6 for power esca lati on testing during initial startup operation.
Catawba 1 Cycle 24 Core Operating Limits Report CNEI-0400-310 Page 20 Revision 0 M H(X , Y) =Margin remaining in core location X , Y relative to Operational DNB limits in the transient power distribution. MAfI(X , Y) is provided in Appendix Table A-3 for normal operation and in Appendix Table A-6 for power escalation testing during initial startup operation. UMR =Uncertainty value for measured radial peaks (UMR = I .0). UMR is set to 1.0 since a factor of I .04 is implicitly included in the variable TILT =Defined in Section 2.6.5. 2.7.3 RRH is defined in Section 2.7.1. 2.7.4 TRH = 0.04 where: TRH = Reduction in 0Ti6.T K 1 setpoint required to compensate for each 1 % that the measured radial peak , F:H (X , Y) exceeds its limit. 2.7.5 F H(X , Y) Penalty Factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2. 2.8 Axial Flux Difference 
-AFD (TS 3.2.3) 2.8.1 Axial Flux Difference (AFD) Limits are provided in Figure 5.
C or e H e i g ht ft 1.05 I.I 0.12 1.8092 1.8553 1.20 1.8 1 02 1.8540 2.40 1.8 0 93 1.8525 3.60 1.8098 1.85 1 4 4.80 1.8097 1.85 1 4 6.00 1.80 97 1.85 1 4 7.20 1.8070 1.8438 8.40 1.8073 1.83 1 9 9.60 1.8 0 72 1.8 10 2 10.80 1.7980 1.7868 11.40 1.7 8 92 1.7652 Catawba 1 C y cle 24 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPs) RFAMARP s Ax ial P ea k 1.2 1.3 J.4 1.5 1.6 1.7 1.8 1.9248 1.9 1 46 1.9 1 79 2.062 1 2.0498 2.0 090 1.9333 1.9248 1.9 1 46 1.9 1 79 2.1 073 2.0 1 9 1 1.9775 1.9009 1.93 1 2 1.9 1 46 1.9 1 79 2.0735 1.9953 1.95 1 9 1.8 76 0 1.9204 1.9 1 46 1.9 1 79 2.0495 1.9656 1.9258 1.8524 1.9058 1.9 1 46 1.9 1 79 2.0059 1.9441 1.9233 1.8538 1.892 1 1.92 1 2 1.9 1 79 1.9336 1.8798 1.8 625 1.8 0 24 1.87 1 6 1.8930 1.8872 1.8723 1.8094 1.7866 1.7332 1.8452 1.857 1 1.8 1 56 I. 7950 1.7359 1.7089 1.6544 1.8093 1.7 9 1 3 1.7375 1.7 1 82 1.6572 1.6347 1.58 08 I. 76 11 1.71 63 1.6538 1.63 1 5 1.5743 1.5573 1.5088 1.7250 1.6645 1.6057 1.5826 1.5289 1.5 0 98 1.4637 1.9 1.8625 1.8306 1.8054 1.7855 1.7 836 1.7 472 1.68 1 2 1.60 I 0 1.530 I 1.4624 1.42 1 8 CNE I-0400-3 10 P age 2 1 R e vi s i o n 0 2.1 3 1.7780 1.3 1 5 1 1.7852 1.3007 I. 7320 1.4633 1.6996 1.4675 1.67 1 4 1.2987 1.6705 1.3293 1.5982 1.287 1 1.5 1 27 1.2 1 82 1.4444 1.1 43 1 1.3 832 I.I 009 1.3458 1.0 670 3.25 1.246 1 1.2235 1.46 1 6 1.3 874 1.2579 1.2602 1.2 1 95 1.1 578 1.09 1 4 1.0470 1.0 1 42  
..... Q) ;i:: 0 '1.. Cii E ..... Q) ...c: f--0 Q) '(;J 0::: '+-< 0 E Q) (.) ..... Q) '1.. CNEI-0400-310 Page 22 Revision 0 Catawba 1 Cycle 24 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits (-20 , 100) (+10 , 100) Unacceptable Operation U nacceptable Operation 90 80 70 Acce ptabl e Operation 60 (-36 , 50) 50 (+21 , 50) 40 30 20 10 40 20 -10 0 10 20 30 40 50 Axial Flux Diff erence (% D e lt a I) NOTE: Compliance with Technical Specification 3.2. l may require more restrictive AFD limits. Refer to the Unit 1 ROD manual for operational AFD limits.
Catawba 1 Cycle 24 Core Operating Limits Report CNEI-0400-310 Pa ge 23 R ev i s ion 0 2.9 Reactor Trip System Instrumentation Setpoints (TS 3.3.1) Table 3.3.1-1
* 2.9.1 Overtemperature LiT Setpoint Parameter Values Parameter Nominal Tavg at RTP Nominal RCS Operating Pressure Overtemperature 
.0.T r eactor tr ip setpoint Overtemperature 
.0.T reactor trip h ea tup set point penalty coefficient Ov e rt em p erat ur e .0.T reactor trip d ep r ess urization setpoint p e nalt y coeffic i e nt Time cons tant s utili zed in the l ead-l ag compe n sator for .0.T Time constant utilized in the l ag compe n sator for .0.T Time co n stants utilized in the lead-lag compensator for T avg Time constant util ized in the measured T avs l ag compensator ft (.0.1) "pos itive" breakpoint ft (.0.1) " n ega ti ve" breakpoint f t (.0.l) "pos itiv e" s l ope ft (.0.1) "negative" s l ope Nominal Value T':S 585.1&deg;f P' = 2235 p s ig K1=1.1978 K2 = 0.03340 1&deg;F K3 = 0.001 60 1 1 psi *1 =8sec. '2 = 3 sec. *3 = 0 sec. <:4 = 22 sec. <: 5 = 4 sec. '6 = 0 sec. = 1 9.0%.0.l =N I A* = 1.769 %.0.T 0 1 %.0.l =N I A* f 1 (61) n egative breakpoints an d s lop es for OT6T are l ess r est ri ctive than the negative bre ak p o int a nd s l ope. T h erefore , during a transient w hi c h challenges th e n egat iv e imb a l ance limit s , OP6T f 2 (6J) limit s wi ll result in a reactor trip before f1 (6 1) limit s are r eac h ed. This m akes impl e m e ntati o n of an OT f 1 (M) n egat i ve breakpoint and s l ope unn ecessary.
CNEI-0400-310 Page 24 Revision 0 Catawba 1 Cycle 24 Core Operating Limits Report 2.9.2 Overpower Setpoint Parameter Values Parameter Nominal Tavg at RTP Overpower reactor trip setpoint Overpower reactor trip penalty Overpower reactor trip heatup setpoint penalty coefficient Time constants utilized in the lead-lag compensator for Time constant utilized in the lag compensator for Time constant utilized in the measured T a v g lag compensator Time constant utilized in the rate-lag controller for T avg "positive" breakpoint fz(M) "negative" breakpoint "positive" slope "negative" slope Nominal Value T":S 585.1&deg;P = 1.0864 K s = 0.02 I 0 P for increasing Tavg K s= 0.00 I 0 P for decreasing Tavg K 6 = 0.001l79/0 P for T > T" K 6 = 0.0 !0 P for T:::; T" 'tt = 8 sec. 't 2 = 3 sec. 't 3 = 0 sec. 't 6 = 0 sec. 't 7 = 10 sec. = 35.0 = -35.0 = 7.0 = 7.0 Catawba 1 Cycle 24 Core Operating Limits Report 2.10 Boron Dilution Mitigation System -BDMS (TS 3.3.9) 2.10.1 Reactor Makeup Water Pump combined flow rate limits: Applicable MODE MOD E 3 MOD E 4 or 5 < 80 gpm < 70 gpm 2.11 RCS Pressure, Temperature and Flow DNB Limits (TS 3.4.1) CNEI-0400-310 Page 25 Revi s ion 0 The RCS pressure , temperature and flow limits for DNB are shown in Table 4. 2.12 Accumulators (TS 3.5.1) 2.12.1 Boron concentration limits during MODES 1 and 2 , and MOD E 3 with RCS pressure > 1000 psi: Parameter Applicabl e Burnup Limit Accumulator minimum boron concentration.
0 -200 E FPD 2 , 500 ppm Accumulator minimum boron concentration. 200.1 -250 E FPD 2 , 500 ppm Accumulator minimum boron concentration. 250.1 -300 EFPD 2 , 466 ppm Accumulator minimum boron concentration. 300.1 -350 E FPD 2 , 349 ppm Accumulator minimum boron concentration. 350.1 -400 EFPD 2 , 260 ppm Accumulator minimum boron concentration. 400.1 -450 EFPD 2 , 186 ppm Accumulator minimum boron concentration. 450.1 -490 EFPD 2 , 111 ppm Accumulator minimum boron concentration. 490.1-531 EFPD 2 , 053 ppm Accumulator minimum boron concentration. 531.1 -541 EFPD 1 , 988 ppm Accumulator maximum boron concentration. 0 -541 EFPD 3 , 075 ppm -I Catawba 1 Cycle 24 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable PARAMETER INDICATION CHANNELS 1. Indicated RCS Average Temperature meter 4 meter 3 computer 4 computer 3 2. Indicated Pressurizer Pressure meter 4 meter 3 computer 4 comp ut er 3 3. RCS Tota l Flow Rate CNEI-0400-310 Page 26 Revision 0 LIMITS :'S 587.2 &deg;F :'S 586.9 &deg;F :'S 587.7 &deg;F :'S 587.5 &deg;F 2: 2209.8 psig 2: 2212.1 psig 2: 2205.8 psig 2: 2207.5 psig 2: 388 , 000 g pm Catawba 1 Cycle 24 Core Operating Limits Report 2.13 Refueling Water Storage Tank -RWST (TS 3.5.4) 2.13.1 Boron concentration limits during MODES l , 2 , 3 , and 4: Parameter R WST minimum boron concentration. R WST maximum boron concentration.
2.14 Spent Fuel Pool Boron Concentration (TS 3.7.15) CNEI-0400-310 Page 27 Revision 0 2 , 700 ppm 3 , 075 ppm 2.14.1 Minimum boron concentration limit for the spent fuel pool. Applicable when fuel assemblies are stored in the spent fuel pool. Parameter Spent fuel pool minimum boron concentration.
2 , 700 ppm 2.15 Refueling Operations 
-Boron Concentration (TS 3.9.1) 2.15.1 Minimum boron concentration limit for filled portions of the Reactor Coolant System , refueling canal , and refueling cavity for MODE 6 conditions.
The minimum boron concentration limit and plant refueling procedures ensure that core K eff remains within the MODE 6 reactivity requirement of K eff :'.S 0.95. Parameter Minimum boron concentration of the Reactor Coolant System , the refueling canal , and the refueling cavity. 2 , 700 ppm CNEI-0400-310 Page 28 Revision 0 Catawba 1 Cycle 24 Core Operating Limits Report 2.16 Standby Shutdown System -(SLC-16.7-9) 2.16.1 Minimum boron concentration limit for the spent fuel pool required for Standby Makeup Pump Water Supply. Applicable for MODES 1 , 2 , and 3. Parameter Spent fuel pool minimum boron concentration for TR 16.7-9-3. Limit 2 , 700 ppm 2.17 Boration Systems Borated Water Source-Shutdown (SLC 16.9-11) 2.17.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature 
:'.S 210&deg;F , and MODES 5 and 6. Parameter Limit NOTE: When cycle burnup is 481 EFPD, Figure 6 may be used to determine the required BAT Minimum Level. BAT minimum boron concentration Volume of 7 , 000 ppm boric acid solution required to maintain SOM at 68&deg;F BAT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-11) R WST minimum boron concentration Volume of 2,700 ppm boric acid solution required to maintain SDM at 68 &deg;F RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-11) 7 , 000 ppm 2 , 000 gallons 13 , 086 gallons (14.9%) 2 , 700 ppm 7 , 000 gallons 48 , 500 gallons (8.7%)
CNEI-0400-310 Page 29 Revision 0 Catawba 1 Cycle 24 Core Operating Limits Report 2.18 Boration Systems Borated Water Source -Operating (SLC 16.9-12) 2.18.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODES l , 2 , and 3 and MODE 4 with all RCS cold leg temperatures
> 210&deg;F*. *NOTE: The SLC 16.9-12 applicability is down to MODE 4 temperatures of > 210&deg;F. The minimum volumes calculated support cooldown to 200&deg;F to satisfy UFSAR Chapter 9 requirements.
Parameter NOTE: When cycle burnup is 2: 481 EFPD, Figure 6 may be used to determine the required BAT Minimum Level. BAT minimum boron concentration Volume of 7 , 000 ppm boric acid solution required to maintain SOM at 210&deg;F BAT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-12) R WST minimum boron concentration Volume of 2 , 700 ppm boric acid solution required to maintain SDM at 210 &deg;F R WST Minimum Shutdown Volume (lncludes the additional volumes listed in SLC 16.9-12) 7 , 000 ppm 13 , 500 gallons 25 , 200 gallons (45.8%) 2 , 700 ppm 57 , 107 gallons 98 , 607 gallons (22.0%)
50.0 45.0 40.0 35.0 30.0 25.0 Q) > Q) _J !;;: 20.0 III 15.0 10.0 5.0 0.0 Catawba 1 Cycle 24 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is;::: 481 EFPD) CNEI-0400-310 Pa ge 30 Revision 0 This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 RCS Boron Concen t ration BAT Level I (ppm) (%l evel) I 0 < 3 0 0 43.0 300 < 500 4 0.0 500 < 700 37.0 700 < 1000 30.0 1000 < 1300 14.9 1300 < 2700 9.8 > 27 0 0 9.8 Unacceptable Operation Acceptable Operation 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 Primary Coolant Boron Concentration (ppmb)
Catawba 1 Cycle 24 Core Operating Limits Report Appendix A Power Distribution Monitoring Factors CNEI-0400-310 Page 31 Revision 0 Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance.
This data was generated in the Catawba I Cycle 24 Maneuvering Analysis calculation file , CNC-1553.05-00-0650.
Due to the size of the monitoring factor data , Appendix A i s controlled electronically within Duke and is not included in the Duke internal copies of the COLR. The Catawba Electrical and Reactor Systems Engineering Section controls monitoring factors via computer files and should be contacted ifthere is a need to access this information.
Appendix A is included in the COLR copy transmitted to the NRC.}}