RA-19-0315, Cycle 26, Core Operating Limits Report, Rev. 1
| ML19228A119 | |
| Person / Time | |
|---|---|
| Site: | Mcguire |
| Issue date: | 07/31/2019 |
| From: | Hager N Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19228A125 | List: |
| References | |
| RA-19-0315 MCEl-0400-368, Rev 1 | |
| Download: ML19228A119 (33) | |
Text
Serial RA-19-0315 McGuire Unit 2, Cycle 26, Core Operating L1m1ts Report, Rev1s1on 1
e.1_.,ouKE
~ ENERGY Facility Code MC Applicable Fac1lit1es Document Number MCEl-0400-368 Document Rev1s1on Number 001 Document EC Number Change Reason CR #02278984 No EC required per AD-NF-ALL-0807, Reload Design Process Document Title MCGUIRE 2 CYCLE 26 CORE OPERATING LIMITS REPOR~
Hager, Nicholas R Ongmator 7/9/2019 Thompson, Ashley Verifier 7/9/2019 S1ry, Steve Safety Analysis DV 7/9/2019 Elkms, Jason R S1te Impact Review 7/9/2019 Blom, Michael A Approver 7/10/2019 Notes
McGuire Umt 2 Cycle 26 Core Operatmg Limits Report Rev1smn 1 July 2019 Calculation Number MCC-1553 05-00-0659, Revision 1 Duke Energy Carolinas, LLC QA Condition 1 MCEI-0400-368 Page 1 Rev1s1on 1 The mformation presented in this report has been prepared and issued m accordance with McGmre Techmcal Specificat10n 5 6 5
McGmre 2 Cycle 26 Core Operatmg Limits Report Implementatmn Instructmns for Rev1s1on 1 Rev1s1on Description and CR Trackmg MCEI-0400-368 Page2 Rev1s10n 1 Rev1s10n 1 of the McGmre Umt 2 Cycle 26 COLR contams hm1ts specific to the reload core This rev1s10n was m1t1ated by CR #02278984 which descnbes the vendor Loss of Coolant Accident (LOCA) analysis update which mcorporated a reqmred surveillance of the m1t1al F.MI assumption mcludmg measurement uncertamty This design cntenon reqmres surveillance of TS 3 2 2 agamst F.MI LOCA hm1t m add1t1on to Loss of Plow Accident (LOFA) Maximum Allowable Radial Peak (MARP) hm1ts Table 3 of the COLR 1s revised to reflect the new hm1ts for this surveillance of the hm1tmg value between LOFA MARPs and F ~H LOCA peakmg hm1ts This rev1s1on also reflects an mcrease m analyzed burnup wmdow to 524 EFPD The McGmre 2 Cycle 26 burnup 1s tracked by NTM #02245502-08 The power d1stnbutlon momtormg factors from Appendix A ofRev1s1on O remam valid and are not transmitted as part of Rev1s1on 1 The 50 59 AR 1s 02224883 Implementation Schedule The McGmre Umt 2 Cycle 26 COLR reqmres the reload 50 59 be approved pnor to 1mplementat1on and fuel loadmg Revision 1 may become effective 1mmed1ately upon receipt The McGmre Umt 2 Cycle 26 COLR will cease to be effective durmg No MODE between cycles 26 and 27 Data FIies to be Implemented No data files are transmitted as part of this document
Revision 0
1 MCEI-0400-368 Page 3 Rev1s10n 1 McGmre 2 Cycle 26 Core Operatmg L1m1ts Report REVISION LOG Effective Date August 2018 July 2019 Pages Affected 1-31, Appendix A*
1-3, 18, 21, 25 COLR M2C26 COLR, Rev 0 M2C26 COLR, Rev 1
- Appendix A contams power d1stnbut1on momtormg factors used m Techmcal Spec1ficat1on Surveillance Appendix A 1s mcluded only m the electromc COLR copy sent to the NRC
McGmre 2 Cycle 26 Core Operatmg Limits Report MCEI-0400-368 Page 4 Rev1s10n 0 1 0 Core Operatmg L1m1ts Report TS Number 2 1 1 3 1 1 3 1 3 314 3 1 5 3 1 5 3 1 6 3 1 6 3 1 8 321 322 323 3 3 1 3 4 1 3 5 1 354 3 7 14 3 9 1 565 This Core Operatmg Limits Report (COLR) has been prepared m accordance with the reqmrements of Techmcal Specification 5 6 5 The Techmcal Specificat10ns that reference this report are hsted below along with the NRC approved analytical methods used to develop and/or determme COLR parameters m Techmcal Specifications NRC Approved COLR Methodology (Section Techmcal Specifications COLR Parameter Sect10n 11 Number)
Reactor Core Safety L1m1ts RCS Temperature and 2 1 6,7,8,9,10,12,15,16,18, Pressure Safety L1m1ts 19 Shutdown Margm Shutdown Margm 22 6,7,8,12,14,15,16,18,l 9 Moderator Temperature Coefficient MTC 23 6,7,8, 14,16, 17 Rod Group Alignment L1m1ts Shutdown Margm 22 6,7,8,12, 14,15,16,18,19 Shutdown Bank Insert10n L1m1ts Shutdown Margm 22 6,7,8,12,14,15, 16,18,19 Shutdown Bank Insertion L1m1ts Shutdown Bank Insertion 24 2,4,6,7,8,9, 10,12, 14, 15, L1m1t 16,18,19 Control Bank Insert10n L1m1ts Shutdown Margm 22 6,7,8,12,14, 15,16,18,19 Control Bank Insert10n L1m1ts Control Bank Insertion 25 2,4,6,7,8,9, 10, 12, 14, 15, L1m1t 16,18,19 Physics Tests Exceptions Shutdown Margm 22 6,7,8, 12, 14, 15, 16, 18, 19 Heat Flux Hot Channel Factor Fq, AFD, OTilTand 26 2,4,6,7,8,9, 10, 12, 15,16, Penalty Factors 18,19 Nuclear Enthalpy Rise Hot Channel FllH,AFDand 27 2,4,6,7,8,9, 10, 12, 15, 16, Factor Penalty Factors 18,19 Axial Flux Difference AFD 28 2,4,6,7,8,15,l6 Reactor Tnp System Instrumentation OT.'lT and OP.'lT 29 6,7,8,9,10,12,15,l6,18, Setpomts Constants 19 RCS Pressure, Temperature, and Flow RCS Pressure, 2 10 6,7,8,9, 10,12,18,19 DNB hm1ts Temperature and Flow Accumulators Max and Mm Boron Cone 2 11 6,7,8,14,16 Refueling Water Storage Tank Max and Mm Boron Cone 2 12 6,7,8,14,16 Spent Fuel Pool Boron Concentration Mm Boron Concentrat10n 2 13 6,7,8,14,16 Refueling Operations - Boron Mm Boron Concentrat10n 2 14 6,7,8,14,16 Concentrat10n Core Operatmg L1m1ts Report (COLR)
Analytical Methods 1 1 None The Selected Licensee Commitments that reference this report are hsted below COLR NRC Approved SLC Selected L1censmg Commitment Section Methodology Number COLR Parameter (Sect10n 1 1 Number) 16 9 14 Borated Water Source-Shutdown Borated Water Volume and 2 15 6,7,8,14,16 Cone forBAT/RWST 16 9 11 Borated Water Source-Operatmg Borated Water Volume and 2 16 6,7,8,14,16 Cone for BA T/R WST 16 9 7 Standby Shutdown System Standby Makeup Pump Water 2 17 6,7,8,14,16 Supply
McGuire 2 Cycle 26 Core Operatmg Limits Report 1 1 Analytical Methods MCEI-0400-368 Page 5 Revision 0 The analytical methods used to determme core operatmg hmrts for parameters 1dent1fied m Techmcal Specifications and previously reviewed and approved by the NRC as specified m Techmcal Spec1ficatton 5 6 5 are as follows 1
WCAP-9272-P-A, "Westmghouse Reload Safety Evaluation Methodology," (W Propnetary)
Revision 0 Report Date July 1985 Not Used 2
WCAP-10054-P-A, "Westmghouse Small Break ECCS Evaluation Model usmg the NOTRUMP Code," (W Propnetary)
Rev1s1on 0 Report Date August 1985 Addendum 2, "Addendum to the Westmghouse Small Break ECCS Evaluation Model Usmg the NOTRUMP Code Safety InJection mto the Broken Loop and COSI Condensation Model," (W Propnetary) (Referenced m Duke Letter DPC-06-101)
Rev1s1on 1 July 1997 3
WCAP-10266-P-A, "The 1981 Version OfWestmghouse Evaluation Model Usmg BASH Code",
(W Propnetary)
Rev1s1on 2 Report Date March 1987 Not Used 4
WCAP-12945-P-A, Volume 1 and Volumes 2-5, "Code Quahfication Document for Best-Estnnate Loss of Coolant Analysis," (W Propnetary)
Rev1s1on Volume 1 (Rev1s1on 2) and Volumes 2-5 (Rev1s1on 1)
Report Date March 1998 5
BA W-10168P-A, "B& W Loss-of-Coolant Accident Evaluat10n Model for Rec1rculatmg Steam Generator Plants," (B&W Propnetary)
Rev1s1on 1 SER Date January 22, 1991 Revision 2 SER Dates August 22, 1996 and November 26, 1996 Rev1s1on 3 SER Date June 15, 1994 Not Used
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report 11 Analytical Methods ( contmued)
MCEI-0400-368 Page 6 Revision 0 6
DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," (DPC Propnetary)
Revision 5a Report Date October 2012 7
DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Propnetary)
Revision 1 Report Date March 2015 8
DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology" Revision 4b Report Date September 2010 9
DPC-NE-2004P-A, "Duke Power Company McGmre and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology usmg VIPRE-01," (DPC Propnetary)
Revision 2a Report Date December 2008 10 DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (DPC Propnetary)
Revision 5 Report Date March 2016 11 DPC-NE-2008P-A, "Fuel Mechamcal Reload Analysis Methodology Usmg TAC03," (DPC Propnetary)
Revision 0 Report Date Apnl 3, 1995 Not Used 12 DPC-NE-2009-P-A, "Westmghouse Fuel Transition Report," (DPC Proprietary)
Revision 3c Report Date March 2017 13 DPC-NE-1004A, "Nuclear Design Methodology Usmg CASM0-3/SIMULATE-3P" Revision la Report Date January 2009 Not Used
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report 11 Analytical Methods (contmued)
MCEI-0400-368 Page 7 Rev1s1on 0 14 DPC-NF-2010-A, "Duke Power Company McGmre Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design "
Rev1s1on 2a Report Date December 2009 15 DPC-NE-2011-PA, "Duke Power Company Nuclear Design Methodology Report for Core Operatmg L1m1ts of Westmghouse Reactors," (DPC Proprietary)
Rev1s10n la Report Date June 2009 16 DPC-NE-1005-PA, "Nuclear Design Methodology Usmg CASM0-4 / SIMULATE-3 MOX,"
(DPC Proprietary)
Rev1s10n I Report Date November 12, 2008 17 DPC-NE-1007-PA, "Cond1tlonal Exempt10n of the EOC MTC Measurement Methodology,"
(DPC and W Proprietary)
Rev1s10n 0 Report Date April 2015 18 WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," (W Proprietary)
Rev1s10n 0 Report Date April 1995 19 WCAP-12610-P-A & CENPD-404-P-A, Addendum I-A, "Optimized ZIRLO'," (W Proprietary)
Rev1s1on 0 Report Date July 2006
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report 2 0 Operatmg L1m1ts MCEI-0400-368 Page 8 Rev1s10n 0 Cycle-specific parameter hmits for the specificat10ns hsted m Sect10n 1 0 are presented m the followmg subsect10ns These hm1ts have been developed usmg the NRC approved methodologies specified m Section 1 1 2 1 Reactor Core Safety Limits (TS 2 11) 2 11 The Reactor Core Safety Limits are shown m Figure 1 2 2 Shutdown Margm - SDM (TS 3 11, TS 3 1 4, TS 3 1 5, TS 3 1 6 and TS 3 1 8) 2 2 1 For TS 3 1 1, SDM shall be::::. 1 3% Af(JK m MODE 2 with k-eff < 1 0 and m MODES 3 and4 2 2 2 For TS 3 1 1, SDM shall be :::::.1 0% Af(JK m MODE 5 2 2 3 For TS 3 1 4, SDM shall be 2::. 1 3% Af(JK m MODES 1 and 2 2 2 4 For TS 3 1 5, SDM shall be 2::. 1 3% Af(JK m MODE 1 and MODE 2 with any control bank not fully mserted 2 2 5 For TS 3 1 6, SDM shall be 2::. 1 3% Af(JK m MODE 1 and MODE 2 with K-eff::::. 1 0 2.2 6 For TS 3 1 8, SDM shall be 2::. 1 3% Af(JK m MODE 2 durmg PHYSICS TESTS
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report Figure 1 Reactor Core Safety L1m1ts Four Loops m Operation MCEI-0400-368 Page9 Rev1s1on 0 DO NOT OPERA TE IN THIS AREA 660 1--~~~~+--~~~~+--~~~~-+
-~~-l 650 640
~ 630 bf)
~
E-;
Cl)
~ 620 610 1945 ps1a 600 ACCEPTABLE OPERATION 00 02 04 06 08 1 0 1 2 Fract10n of Rated Thermal Power
McGmre 2 Cycle 26 Core Operatmg Limits Report 2 3 Moderator Temperature Coefficient - MTC (TS 3 13) 2 3 1 The Moderator Temperature Coefficient (MTC) L1m1ts are MCEI-0400-368 Page 10 Rev1s10n 0 MTC shall be less pos1t1ve than the upper limits shown m Figure 2 BOC, ARO, HZP MTC shall be less positive than O 7E-04 Af0'K/°F EOC, ARO, RTP MTC shall be less negative than the -4 3E-04 Af0'K/°F lower MTC hmit 2 3 2 300 PPM MTC Surveillance Limit is Measured 300 PPM ARO, eqmhbnum RTP MTC shall be less negative than or equal to -3 65E-04 Af</K/°F 2 3 3 The Revised Predicted near-EOC 300 PPM ARO RTP MTC shall be calculated usmg the procedure contamed m DPC-NE-1007-PA If the Revised Predicted MTC is less negative than or equal to the 300 PPM SR 3 1 3 2 Surveillance Limit, and all benchmark data contamed m the surveillance procedure is satisfied, then a MTC measurement m accordance with SR 3 1 3 2 1s not reqmred to be performed 2 3 4 60 PPM MTC Surveillance Limit is 60 PPM ARO, eqmhbnum RTP MTC shall be less negative than or equal to
-4 125E-04 Af(/K/°F Where BOC = Begmnmg of Cycle (burnup correspondmg to the most positive MTC)
EOC = End of Cycle ARO = All Rods Out HZP = Hot Zero Power RTP = Rated Thermal Power PPM = Parts per m1lhon (Boron) 2 4 Shutdown Bank Insertion L1m1t (TS 3 1 5) 2 4 1 Each shutdown bank shall be withdrawn to at least 222 steps Shutdown banks are withdrawn m sequence and with no overlap 2 5 Control Bank Insertion L1m1ts (TS 3 1 6) 2 5 1 Control banks shall be w1thm the msert1on, sequence, and overlap limits shown m Figure 3 Specific control bank withdrawal and overlap limits as a function of the fully withdrawn pos1t1on are shown m Table 1
10 09 0.8
=
Cl,)...
07 Cj
~
Cl,) O_,
u~ 06 Cl,)
0 i~ 05 t <l 04
~',:f' 59
~~ 03 i.......
o-
~
02 I-,
Cl,)
"t:I 0
01
~
00 McGmre 2 Cycle 26 Core Operatmg Lmuts Report Figure 2 MCEI-0400-368 Page 11 Rev1s10n 0 Moderator Temperature Coefficient Upper Limit Versus Power Level Unacceptable Operation Acceptable Operation 0
10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power NOTE Compliance with Techmcal Specification 3 1 3 may reqmre rod withdrawal limits Refer to OP/2/A/6100/22 Umt 2 Data Book for details
231 220 200
= 180
~ =
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report Figure 3 MCEI-0400-3 68 Page 12 Rev1s1on 0 Control Bank Insertion Limits Versus Percent Rated Thermal Power Fully Withdrawn (Maximum= 23 l~
/
/
/
/
/
/
~
/
/
V
/
Fully Withdrawn
/
I/
/
Control Bank B -
(Mm1mum = 222) 1,,,-
/
I (100% 161) r=
/
/
]
160 H co%, 163)
/
~ 140 I/
/
/
/
/
C.
/
/
Ill
~ 120 1 Control Bank C 1
/
/
/
=
Si....
0
~ =
0
-e Ill "' = -
"O 0
~
/
/
/
100
~
/
/
./
/
80
/
/
/
I/
/
Control Bank D 1
/
60
/
/
/
/
40 ~ (0% 47)
/
/
/
20
~
~Fully Inserted
/
/
(30% 0) I _,
0 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), Control Bank C (CC), and Control Bank B (CB) can be calculated by Bank CD RIL = 2 3(P)- 69 {30 < P < 100}
Bank CC RIL = 2 3(P) +47
{O < P < 76 1) for CC RIL = 222 {76 1 < P < 100}
Bank CB RIL = 2 3(P) + 163 {O < P:::.. 25 7} for CB RJL = 222 {25 7 < P < 100}
where P = %Rated Thermal Power NOTE Compliance with Technical Specification 3 1 3 may reqmre rod withdrawal limits Refer to OP/2/A/6100/22 Umt 2 Data Book for details
MCEI-0400-368 Page 13 Rev1s1on 0 McGmre 2 Cycle 26 Core Operatmg L1m1ts Report Table 1 RCCA Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Fully Withdrawn at 223 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB Banke BankD 0 Start 0
0 0
0 Start 0
0 0
116 0 Start 0
0 116 0 Start 0
0 222 Stop 106 0
0 223 Stop 107 0
0 222 116 0 Start 0
223 116 0 Start 0
222 222 Stop 106 0
223 223 Stop 107 0
222 222 116 0 Start 223 223 116 0 Start 222 222 222 Stop 106 223 223 223 Stop 107 Fully Withdrawn at 224 Steps Fully Withdrawn at 225 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB Banke BankD 0 Start 0
0 0
0 Start 0
0 0
116 0 Start 0
0 116 0 Start 0
0 224 Stop 108 0
0 225 Stop 109 0
0 224 116 0 Start 0
225 116 0 Start 0
224 224 Stop 108 0
225 225 Stop 109 0
224 224 116 O Start 225 225 116 0 Start 224 224 224 Stop 108 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Fnlly Withdrawn at 227 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB BankC BankD 0 Start 0
0 0
O Start 0
0 0
116 0 Start 0
0 116 0 Start 0
0 226 Stop 110 0
0 227 Stop 111 0
0 226 116 0 Start 0
227 116 0 Start 0
226 226 Stop 110 0
227 227 Stop Ill 0
226 226 116 0 Start 227 227 116 0 Start 226 226 226 Stop 110 227 227 227 Stop Ill Fully Withdrawn at 228 Steps Fully Withdrawn at 229 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB Banke BankD 0 Start 0
0 0
0 Start 0
0 0
116 0 Start 0
0 116 0 Start 0
0 228 Stop 112 0
0 229 Stop 113 0
0 228 116 0 Start 0
229 116 0 Start 0
228 228 Stop 112 0
229 229 Stop 113 0
228 228 116 0 Start 229 229 116 0 Start 228 228 228 Stop 112 229 229 229 Stop 113 Fully Withdrawn at 230 Steps Fully Withdrawn at 231 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB BankC BankD 0 Start 0
0 0
0 Start 0
0 0
116 0 Start 0
0 116 0 Start 0
0 230 Stop 114 0
0 231 Stop 115 0
0 230 116 0 Start 0
231 116 0 Start 0
230 230 Stop 114 0
231 231 Stop 115 0
230 230 116 0 Start 231 231 116 0 Start 230 230 230 Stop 114 231 231 231 Stop 115
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report 2 6 Heat Flux Hot Channel Factor - FQ(X,Y,Z) (TS 3 21)
MCEI-0400-368 Page 14 Rev1s10n 0 2 6 1 FQ(X,Y,Z) steady-state hm1ts are defined by the followmg relat1onsh1ps
- where, F RTP *K(Z)IP Q
F ~TP *K(Z)/0 5 for P > 0 5 for P ~ 0 5 P = (Thermal Power)/(Rated Power)
Note The measured FQ(X,Y,Z) shall be mcreased by 3% to account for manufacturmg tolerances and 5% to account for measurement uncertamty when comparmg agamst the LCO hm1ts The manufacturmg tolerance and measurement uncertamty are 1mphc1tly mcluded m the FQ surveillance hm1ts as defined m Sect10ns 2 6 5 and 2 6 6 2 6 2 F RTP = 2 70 x K(BU)
Q 2 6 3 K(Z) 1s the normalized FQ(X,Y,Z) as a function of core height The K(Z) function for Westmghouse RF A fuel 1s provided m Figure 4 2 6 4 K(BU) 1s the normalized FQ(X,Y,Z) as a function of burnup F ~TP with the K(BU) penalty for Westmghouse RF A fuel 1s analytically confirmed m cycle-specific reload calculations K(BU) 1s set to 1 0 at all burnups The followmg parameters are reqmred for core momtormg per the Surveillance Reqmrements of Techmcal Spec1ficat1on 3 2 1 L
Fg(X,Y,Z)
- M0(X,Y,Z) 2 6 5 FQ(X,Y,Z)OP =
UMT *MT* TILT
where McGmre 2 Cycle 26 Core Operatmg Limits Report MCEI-0400-368 Page 15 Rev1s1on 0 Ff (X,Y,Z)OP =
Cycle dependent maximum allowable design peakmg factor that ensures FQ(X,Y,Z) LOCA limit will be preserved for operation w1thm the LCO hm1ts FQL (X,Y,z)OP mcludes allowances for calculation and measurement uncertamties Frf (X,Y,Z) = Design power distnbution for FQ Frf (X,Y,Z) 1s provided m Appendix Table A-1 for normal operatmg conditions, and m Appendix Table A-4 for power escalation testmg durmg m1tJal startup operation MQ(X,Y,Z) = Margm remammg m core location X,Y,Z to the LOCA limit m the transient power distnbutJon MQ(X,Y,Z) 1s provided m Appendix Table A-1 for normal operatmg conditions and m Appendix Table A-4 for power escalation testmg durmg mitial startup operation UMT = Total Peak Measurement Uncertamty (UMT = 1 05)
MT = Engmeermg Hot Channel Factor (MT= 1 03)
TILT Peakmg penalty to account for allowable quadrant power tilt ratio of 1 02 (TILT= 1 035) 2 6 6 Ftcx,Y,z)RPs =
F~(X, Y,Z)
- Mc(X, Y,Z)
UMT *MT* TILT where Ft(X,Y,Z)RPS =
Cycle dependent maximum allowable design peakmg factor that ensures the FQ(X,Y,Z) Centerlme Fuel Melt (CFM) hm1t will be preserved for operation w1thm the LCO limits L
FQ(X,Y,Z)RPS mcludes allowances for calculation and measurement uncertamt1es D
FQ(X,Y,Z) = Defined m Sect10n 2 6 5
McGmre 2 Cycle 26 Core Operatmg Limits Report MCEI-0400-368 Page 16 Rev1s1on 0 Mc(X,Y,Z) = Margm remammg to the CFM hmit m core location X,Y,Z from the transient power distribution Mc(X,Y,Z) is provided m Appendix Table A-2 for normal operatmg conditions and m Appendix Table A-5 for power escalation testmg durmg mitial startup operat10n UMT = Defined m Section 2 6 5 MT = Defined m Section 2 6 5 TILT = Defined m Section 2 6 5 2 6 7 KSLOPE = 0 0725 where KSLOPE is the adjustment to K1 value from the OT~T tnp setpomt reqmred to compensate for each 1 % that F(: (X,Y,Z) exceeds Ft (X,Y,Z)RPs 2 6 8 FQ(X,Y,Z) penalty factors for Techmcal Specification Surveillances 3 2 1 2 and 3 2 1 3 are provided m Table 2
McGmre 2 Cycle 26 Core Operatmg Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westmghouse RF A Fuel MCEI-0400-368 Page 17 Rev1s1on 0 1200 ~--------------------------~
0 800
§: 0 600
~
0 400 0 200 0 000 (0 0, I 00)
(4 0, I 00)
(4 0, 0 9259)
Core Height (ft)
K(Z) 00 I 0
- 5 4 0 I 0
>40 0 9259 12 0 0 9259 00 20 40 60 Core Height (ft)
(12 0, 0 9259) 80 10 0 12 0
McGmre 2 Cycle 26 Core Operatmg Limits Report Table2 FQ(X,Y,Z) and F.AfI(X,Y) Penalty Factors MCEl-0400-368 Page 18 Rev1s10n 1 For Techmcal Specification Surveillance's 3 2 1 2, 3 2 1 3 and 3 2 2 2 Burnup FQ(X,Y,Z)
F,rn(X,Y)
(EFPD)
Penalty Factor (%)
Penalty Factor(%)
4 2 00 2 00 12 2 00 2 00 25 2 00 2 00 50 2 00 2 00 75 2 00 2 00 100 2 00 2 00 125 2 00 2 00 150 2 00 2 00 175 2 00 2 00 200 2 00 2 00 225 2 00 2 00 250 2 00 2 00 275 2 00 2 00 300 2 00 2 00 325 2 00 2 00 350 2 00 2 00 375 2 00 2 00 400 2 00 2 00 425 2 00 2 00 450 2 00 2 00 460 2 00 2 00 475 2 00 2 00 489 2 00 2 00 494 2 00 2 00 499 2 00 2 00 509 2 00 2 00 524 2 00 2 00 Note Lmear mterpolatron rs adequate for mtermedrate cycle bumups All cycle burn ups outside of the range of the table shall use a 2% penalty factor for both FQ(X,Y,Z) and F AfI(X,Y) for compliance with the Techmcal Spec1ficat1on Surveillances 3 2 1 2, 3 2 1 3 and 3 2 2 2
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report MCEI-0400-368 Page 19 Rev1s1on 0 2 7 Nuclear Enthalpy Rise Hot Channel Factor - F m(X,Y) (TS 3 2 2)
F Af-I steady-state limits referred tom Techmcal Specification 3 2 2 1s defined by the followmg relationship where F~ (X, Ylco 1s the steady-state, maximum allowed radial peak and mcludes allowances for calculation/measurement uncertamty MARP(X, Y) = Cycle-specific operatmg hmit Maximum Allowable Radial Peaks MARP(X,Y) radial peakmg limits are provided m Table 3 Thermal Power p = Rated Thermal Power RRH = Thermal Power reduct10n reqmred to compensate for each 1 % that the measured radial peak, F~ (X, Y), exceeds its hmit RRH also is used to scale the MARP limits as a function of power per the FkH (X, Y)Lco equat10n (RRH = 3 34 (0 0 < P:::: 1 0))
The followmg parameters are reqmred for core momtormg per the surveillance reqmrements of Techmcal Spec1ficat10n 3 2 2 272 pL (X Y)SURV = F~H(X, Y) X ML\\H(X, Y) t.H UMRxTILT where F~H (X,Y)SURV = C l d d
ll bl d k
.c:
ye e epen ent maximum a owa e esign pea mg 1actor that ensures the F L\\ttCX, Y) hmit will be preserved for operation withm the LCO hmits pL (X Y)suRv mcludes allowances for llH calculat10n/measurement uncertamty
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report MCEl-0400-368 Page 20 Revision 0 D
D F Afl (X,Y) = Design radial power d1stnbut1on for F LiH F LiH (X,Y) 1s provided m Appendix Table A-3 for normal operation and m Appendix Table A-6 for power escalation testmg durmg m1tial startup operat10n M~H(X,Y) = The margm remammg m core location X,Y relative to the Operational DNB hm1ts m the transient power d1stnbut1on Mm(X,Y) 1s provided m Appendix Table A-3 for normal operation and m Appendix Table A-6 for power escalation testmg durmg 1mt1al startup operation UMR = Uncertamty value for measured radial peaks (UMR = 1 0)
UMR 1s 1 0 smce a factor of 1 04 1s 1mphc1tly mcluded m the variable MiiH(X,Y)
TILT = Peakmg penalty to account for allowable quadrant power tilt rat10 of 1 02 (TILT= 1 035) 2 7 3 RRH 1s defined m Sect10n 2 7 1 2 7 4 TRH= 0 04 where TRH = Reduct10n m the OT~ T K 1 setpomt reqmred to compensate for each 1 %
that the measured radial peak, F~ (X,Y) exceeds its hm1t 2 7 5 FliH (X,Y) penalty factors for Techmcal Spec1:ficatlon Surveillance 3 2 2 2 are provided m Table 2 2 8 Axial Flux Difference - AFD (TS 3 2 3) 2 8 1 The Axial Flux Difference (AFD) L1m1ts are provided m Figure 5
MCEI-0400-368 Page 21 Rev1s1on 1 McGmre 2 Cycle 26 Core Operatmg Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS)
RFA Steady State Limltmg Value Between Loss of Flow Accident (LOFA) MARPs and FMlwcA Core Axial Peak Height ft 1 05 11 1 2 1 3 1 4 1 5 1 6 1 7 1 8 1 9 21 3
3 25 0 12 I 6058 I 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 3151 1 2461 1 20 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 3007 1 2235 2 40 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 4633 1 4616 3 60 1 6058 I 6058 1 6058 1 6058 l 6058 I 6058 1 6058 I 6058 1 6058 1 6058 1 6058 1 4675 1 3874 4 80 1 6058 1 6058 1 6058 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 1 6058 I 2987 1 2579 6 00 1 6058 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 1 6058 I 6058 1 6058 I 6058 I 3293 I 2602 7 20 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 I 6058 1 6058 I 5982 I 2871 I 2195 8 40 I 6058 I 6058 I 6058 1 6058 1 6058 I 6058 I 6058 I 6058 I 6058 1 6010 I 5127 I 2182 1 1578 9 60 I 6058 I 6058 1 6058 1 6058 I 6058 I 6058 1 6058 1 6058 I 5808 I 5301 I 4444 I 143 I I 0914 10 80 I 6058 I 6058 I 6058 1 6058 1 6058 1 6058 1 5743 1 5573 1 5088 1 4624 1 3832 1 1009 1 0470 1140 I 6058 1 6058 1 6058 1 6058 1 6057 1 5826 1 5289 1 5098 1 4637 1 4218 1 3458 1 0670 1 0142
I.
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~ =
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~
-=
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~
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..... =
C
~.,
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~
,:lo;
-50 McGmre 2 Cycle 26 Core Operatmg L1m1ts Report Figure 5 MCEI-0400-368 Page 22 Rev1s1on 0 Percent of Rated Thermal Power Versus Percent Axial Flux Difference L1m1ts
(-18, 100)
(+10, 100)
Unacceptable Operat10n 90 Unacceptable Operat10n 80 Acceptable Operat10n 70 60 50
(-36, 50)
(+21, 50) 40 30 20 10
-40
-30
-20
-10 0
10 20 30 40 50 Axial Flux Difference (% Delta I)
NOTE Compliance with Techmcal Spec1ficat10n 3 2 1 may reqmre more restnctive AFD hm1ts Refer to OP/2/A/6100/22 Umt 2 Data Book for more details
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report MCEI-0400-368 Page 23 Revision 0 2 9 Reactor Trip System Instrumentation Setpomts (TS 3 3 1) Table 3 3 1-1 2 91 Overtemperature AT Setpomt Parameter Values Parameter Nommal Tavg at RTP Nommal RCS Operatmg Pressure Overtemperature ~ T reactor tnp setpomt Overtemperature ~ T reactor tnp heatup setpomt penalty coefficient Overtemperature ~ T reactor tnp depressunzatlon setpomt penalty coefficient Time constants utilized m the lead-lag compensator for ~T Time constant utilized m the lag compensator for ~ T Time constants utilized m the lead-lag compensator for Tavg Time constant utilized m the measured T avg lag compensator f1 (M) "positive" breakpomt f 1 (M) "negative" breakpomt f1 (M) "positive" slope f1 (~I) "negative" slope T ~585 l°F P = 2235 ps1g K1 ~ 1 1978 K2 = 0 03341°F K3 = 0 001601/psi
- 1 ~ 8 sec
- 2 ~ 3 sec "3 ~ 2 sec
- 4 ~ 28 sec
- 5 ~ 4 sec
- 6 ~ 2 sec
= 19 0 %~I
=NIA*
= 1 769 %~Toi %M
=NIA*
The f1 (~I) "negative" breakpomt and the f1 (~I) "negative" slope are less restnct1ve than the OP~T f2(~I) negative breakpomt and slope Therefore, dunng a transient which challenges the negative imbalance hm1ts, the OP~T f2(M) limits will result ma reactor tnp before the OT~T f1 (~I) limits are reached This makes 1mplementat1on of the OT~T f1 (M) negative breakpomt and slope unnecessary
MCEI-0400-368 Page 24 Rev1s10n 0 McGmre 2 Cycle 26 Core Operatmg Limits Report 2 9 2 Overpower dT Setpomt Parameter Values Parameter Nominal Tavg at RTP Overpower L~.T reactor tnp setpoint Overpower ~ T reactor tnp Penalty Overpower ~ T reactor tnp heatup setpoint penalty coefficient Time constants utilized in the lead-lag compensator for ~ T Time constant utilized in the lag compensator for ~ T Time constant utilized in the measured T avg lag compensator Time constant utilized in the rate-lag controller for T avg fz(Af) "positive" breakpoint fz(Af) "negative" breakpoint fz(Af) "positive" slope fz(Af) "negative" slope T" :S 585 1 °F Ki :S 1 0864 Ks = 0 02/°F for increasing Tavg Ks = 0 0 for decreasing Tavg K6 = 0 001179/°F for T > T' K6 = 0 0 for T :S T '
'tl 2:_ 8 sec
't2 :S 3 sec
= 35 0 %~I
= -35 0 %AI
= 7 0 %~Toi %~I
= 7 0 %~Toi %~I
McGmre 2 Cycle 26 Core Operatmg Limits Report 2 10 RCS Pressure, Temperature and Flow Limits for DNB (TS 3 4 1)
MCEI-0400-368 Page 25 Rev1s10n 1 2 10 1 RCS pressure, temperature and flow hmits for DNB are shown m Table 4 2 11 Accumulators (TS 3 5 1) 2 111 Boron concentration hmits durmg MODES 1 and 2, and MODE 3 with RCS pressure > 1000 psi Parameter Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator mm1mum boron concentration Accumulator maximum boron concentration Applicable Bumup 0-200 EFPD 200 1 - 250 EFPD 250 1 - 300 EFPD 300 1 - 350 EFPD 350 1 - 400 EFPD 400 1 - 450 EFPD 450 1 - 494 EFPD 494 1 - 509 EFPD 509 1 - 524 EFPD 0- 524 EFPD 2 12 Refuelmg Water Storage Tank - RWST (TS 3 5 4) 2 12 1 Boron concentration limits durmg MODES 1, 2, 3, and 4 Parameter R WST mmimum boron concentration R WST maximum boron concentration 2,475 ppm 2,475 ppm 2,475 ppm 2,406 ppm 2,290 ppm 2,215 ppm 2,144 ppm 2,074 ppm 2,047 ppm 2,875 ppm 2,675 ppm 2,875 ppm
1 McGmre 2 Cycle 26 Core Operatmg L1m1ts Report Table 4 Reactor Coolant System DNB Parameters No Operable Parameter Ind1cat10n Channels Indicated RCS Average Temperature meter 4
meter 3
computer 4
computer 3
2 Indicated Pressunzer Pressure meter 4
meter 3
computer 4
computer 3
3 RCS Total Flow Rate MCEI-0400-368 Page 26 Rev1s1on 0 Limits
- S 587 2 °F
- S 586 9 °F
- S 587 7 °F
=::587 5 °F
- _2212 3 psig
- 2215 0 psig
- 2209 1 psig
- 2211 3 psig
- 388,000 gpm
McGmre 2 Cycle 26 Core Operatmg Limits Report 2 13 Spent Fuel Pool Boron Concentration (TS 3 7 14)
MCEI-0400-368 Page 27 Rev1s1on 0 2 13 1 Mmimum boron concentration hmit for the spent fuel pool Apphcable when fuel assembhes are stored m the spent fuel pool Parameter Spent fuel pool mm1mum boron concentration 2,675 ppm 2 14 Refuelmg Operations - Boron Concentration (TS 3 9 1) 2 14 1 Mmimum boron concentrat10n hmit for the filled portions of the Reactor Coolant System, refuelmg canal, and refuelmg cavity for MODE 6 conditions The mmimum boron concentration hmit and plant refuelmg procedures ensure that core Keff remams withm MODE 6 reactivity reqmrement of Keff::: 0 95 Parameter Mm1mum boron concentration of the Reactor Coolant System, the refuelmg canal, and the refuelmg cavity 2,675 ppm
MCEI-0400-368 Page 28 Rev1s1on 0 McGmre 2 Cycle 26 Core Operatmg Limits Report 2 15 Borated Water Source-Shutdown (SLC 16 9 14) 2 15 1 Volume and boron concentrations for the Bone Acid Tank (BAT) and the Refuelmg Water Storage Tank (RWST) durmg MODE 4 with any RCS cold leg temperature :S: 300 °F and MODES 5 and 6 Parameter Note When cycle burnup 1s > 464 EFPD, Figure 6 may be used to determme reqmred BAT mm1mum level BAT mmimum contamed borated water volume BAT mmimum boron concentration BAT mmimum water volume reqmred to mamtam SDM at 7,150 ppm R WST mmimum contamed borated water volume R WST mmimum boron concentration R WST mm1mum water volume reqmred to mamtam SDM at 2,675 ppm 10,599 gallons 13 6% Level 7,150 ppm 2,300 gallons 47,700 gallons 41 mches 2,675 ppm 8,200 gallons
MCEI-0400-368 Page 29 Rev1s1on 0 McGmre 2 Cycle 26 Core Operatmg Limits Report 2 16 Borated Water Source - Operatmg (SLC 16 9 11) 2 16 1 Volume and boron concentrations for the Bone Acid Tank (BAT) and the Refuelmg Water Storage Tank (RWST) durmg MODES 1, 2, 3, and MODE 4 with all RCS cold leg temperature> 300 °F *
- Note The SLC 16 9 11 apphcab1hty 1s down to Mode 4 temperatures of > 300°F The mm1mum volumes calculated support cooldown to 200°F to satis UFSAR Cha ter 9 re mrements Parameter Note When cycle burnup 1s > 464 EFPD, Figure 6 may be used to determme reqmred BAT m1mmum level BAT mmimum contamed borated water volume BAT mmimum boron concentration BAT mmimum water volume reqmred to mamtam SDM at 7,150 ppm R WST mmimum contamed borated water volume RWST mmimum boron concentration RWST maximum boron concentration (TS 3 5 4)
RWST mmimum water volume reqmred to mamtam SDM at 2,675 ppm 2 17 Standby Shutdown System - (SLC-16 9 7) 22,049 gallons 38 0% Level 7,150 ppm 13,750 gallons 96,607 gallons 103 6 mches 2,675 ppm 2,875 ppm 5 7, 107 gallons 2 17 1 Mmimum boron concentration hmit for the spent fuel pool reqmred for Standby Makeup Pump Water Supply Apphcable for MODES 1, 2, and 3 Parameter Spent fuel pool mmimum boron concentration for TR 16 9 7 2 2,675 ppm
McGmre 2 Cycle 26 Core Operatmg Limits Report Figure 6 Bone Acid Storage Tank Indicated Level Versus RCS Boron Concentration (Vahd When Cycle Burnup 1s > 464 EFPD)
MCEI-0400-3 68 Page 30 Rev1s1on 0 This figure mcludes additional volumes hsted m SLC 16 9 14 and 16 911 400,-------------,-------------------,-1----------,I,---------------,
j i
l I
I I
35 0 30 0 10 0 1
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-~ - - I I
I L
T I I r
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Acceptable J
- -+-
i I
I Unacceptable Operation I RCS Boron I Concentration, BAT Level (ppm)
I (%level)
I 37 o 0 < 300 300 < 500 33 0 500 <7~
280 1000 < 1300 13 6
> 1300 8 7 I
I I
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1--~-- ; __ _J_ ~--J~~ l I
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~--i ~ --'----- : --~--i - '"-r --+---
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l j
I OO+-----i-l __ ~l __ ~l __ ~lr-_~--,---~--,----i-1 __
~1 __ ~'------r---rl -----1 0
200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 RCS Boron Concentrat10n (ppmb)
McGmre 2 Cycle 26 Core Operatmg L1m1ts Report MCEI-0400-368 Page 31 Rev1s10n 0 NOTE Appendix A contams power d1stnbut10n momtormg factors used m Techmcal Specification Surveillance This data was generated m the McGmre 2 Cycle 26 Maneuvering Analysis calculation file, MCC-1553 05-00-0652 Due to the size of the momtormg factor data, Appendix A 1s controlled electromcally w1thm the Duke document management system and 1s not mcluded m the Duke mternal copies of the COLR The Plant Reactor Engmeermg and Support Systems section will control this mformat1on via computer file(s) and should be contacted 1f there 1s a need to access this information Appendix A 1s mcluded m the COLR copy transmitted to the NRC