ML19228A118
| ML19228A118 | |
| 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 MCEI-0400-379, Rev 1 | |
| Download: ML19228A118 (33) | |
Text
Serial RA-19-0315 McGuire Unit 1, Cycle 27, Core Operating L1m1ts Report, Rev1s1on 1
({_.,DUKE
~ ENERGY Facility Code Applicable Fac1ht1es Document Number Document Rev1s1on Number Document EC Number Change Reason Document Title Hager, Nicholas R Thompson, Ashley Phelps, Timothy P Petrea, Rebecca L Blom, Michael A Notes MC MC MCEl-0400-379 001 CR #02278984 No EC required per AD-NF-ALL-0807, Reload Design Process McGuire 1 Cycle 27 Core Operating L1m1ts Report Originator 7/8/2019 Verifier 7/8/2019 Concurrent Verifier 7/8/2019 S1te Impact Review 7/8/2019 Approver 7/8/2019
McGmre Umt 1 Cycle 27 Core Operatmg Limits Report Revision 1 July 2019 Calculation Number MCC-1553 05-00-0668, Revision 1 Duke Energy Carolmas, LLC QA Condition 1 MCEI-0400-379 Page 1 Rev1s10n 1 The mformat10n presented m this report has been prepared and issued m accordance with McGmre Techmcal Spec1ficat10n 5 6 5
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report Implementation Instruct1ons For Rev1s10n 1 Rev1s10n Description and CR Trackmg MCEI-0400-379 Page2 Rev1s10n 1 Revision 1 of the McGmre Umt 1 Cycle 27 COLR contams limits specific to the reload core This revision was imtiated by CR #02278984 which descnbes the vendor Loss of Coolant Accident (LOCA) analysis update which mcorporated a reqmred surveillance of the m1t1al F LllI assumption mcludmg measurement uncertamty This design cntenon reqmres surveillance of TS 3 2 2 agamst F LllI LOCA hm1t m add1t1on to Loss of Flow Accident (LOF A) Maximum Allowable Radial Peak (MARP) hm1ts Table 3 of the COLR 1s revised to reflect the new limits for this surveillance of the hm1tmg value between LOFA MARPs and F LllI LOCA peakmg limits This rev1s1on also issues an adjustment to the burnup dependent mm1mum reqmred CLA boron concentration for burnups > 200 EFPD The revised values reflect the actual Cycle N-1 shutdown cond1t10ns, as opposed to applymg an assumed shutdown wmdow The power d1stnbut1on momtormg factors from Appendix A of Rev1s10n O remam vahd and are not transmitted as part of Rev1s1on 1 Implementation Schedule The McGmre Umt 1 Cycle 27 COLR reqmres the reload 50 59 (AR #02258276) be approved pnor to 1mplementat10n and fuel loadmg Revis10n 1 may become effective immediately upon receipt The McGmre Umt 1 Cycle 27 COLR will cease to be effective dunng No MODE between cycles 27 and 28 Data Files to be Implemented No data files are transmitted as part ofth1s document
Rev1s1on 0
1 MCEI-0400-379 Page 3 Rev1s10n 1 McGmre 1 Cycle 27 Core Operatmg Limits Report Effective Date March 2019 July 2019 REVISION LOG Pages Affected 1-31, Appendix A*
1-3, 21, 25 COLR M1C27 COLR, Rev 0 M1C27 COLR, Rev 1
- Appendix A contams power d1stnbutlon momtormg factors used m Technical Specification Surveillance Appendix A 1s mcluded only m the electromc COLR copy sent to the NRC
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report MCEI-0400-379 Page4 Revision 0 1 0 Core Operatmg Limits Report TS Number 2 1 1 3 1 I 3 1 3 3 I 4 3 I 5 3 1 6 3 I 8 3 2 I 322 323 3 3 1 3 4 I 3 5 I 354 3 7 14 3 9 I 565 This Core Operatmg Limits Report (COLR) has been prepared m accordance with the reqmrements ofTechmcal Specification 5 6 5 The Techmcal Specifications that reference this report are listed below along with the NRC approved analytical methods used to develop and/or determme COLR parameters m Techmcal Specificat10ns NRC Approved COLR Methodology (Section Techmcal Specifications COLR Parameter Section 11 Number)
Reactor Core Safety L1m1ts RCS Temperature and Pressure 2 1 6,7,8,9,10,12,15,16, 18, Safety L1m1ts 19 Shutdown Margm Shutdown Margm 22 6,7,8,12,14, 15,16, 18,19 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 Insertion L1m1ts Shutdown Margm 22 2,4,6, 7,8,9, I 0, 12, 14,15, Shutdown Bank Insert10n L1m1t 24 16,18,19 Control Bank Insertion L1m1ts Shutdown Margm 22 2,4,6,7,8,9, I 0, 12, 14, 15, Control Bank Insert10n L1m1t 25 16,18,19 Physics Tests Except10ns Shutdown Margm 22 6,7,8, 12,14,15,16,18,19 Heat Flux Hot Channel Factor FQ 26 2,4,6,7,8,9,l 0, 12, 15, 16, AFD 26 18,19 OT~T 29 Penalty Factors 26 Nuclear Enthalpy Rise Hot Channel Fm 27 2,4,6,7,8,9,10, 12, 15,16, Factor Penalty Factors 27 18,19 Axial Flux Difference AFD 28 2,4,6,7,8,15,l6 Reactor Tnp System Instrumentat10n OT~T 29 6, 7,8,9, 10, 12, 15, 16, 18, Setpomts OP~T 29 19 RCS Pressure, Temperature, and RCS Pressure, Temperature and 2 10 6,7,8,9,10, 12, 18, 19 Flow DNB limits Flow Accumulators Max and Mm Boron Cone 2 11 6,7,8,14,16 Refuelmg Water Storage Tank Max and Mm Boron Cone 2 12 6,7,8,14,16 Spent Fuel Pool Boron Concentration Mm Boron Concentration 2 13 6,7,8,14,16 Refueling Operations - Boron Mm Boron Concentrat10n 2 14 6,7,8,14,16 Concentration Core Operatmg L1m1ts Report Analytical Methods I I None (COLR)
The Selected Licensee Commitments that reference this report are hsted below SLC COLR NRC Approved Selected Lrcensmg Commitment Section Methodology Number COLR Parameter (Section 1 1 Number) 16 9 14 Borated Water Source - Shutdown Borated Water Volume and 2 15 6,7,8,14,16 Cone for BAT/RWST 16 9 11 Borated Water Source - Operatmg Borated Water Volume and 2 16 6,7,8,14,16 Cone for BAT/RWST 16 9 7 Standby Shutdown System Standby Makeup Pump Water 2 17 6,7,8,14,16 Suooly
McGmre 1 Cycle 27 Core Operatmg Limits Report 1 1 Analytical Methods MCEI-0400-379 Page 5 Rev1s10n 0 The analytical methods used to determme core operatmg hm1ts for parameters identified m Techmcal Specifications and prevrously reviewed and approved by the NRC as specified m Techmcal Specification 5 6 5 are as follows 1
WCAP-9272-P-A, "Westmghouse Reload Safety Evaluation Methodology," (W Propnetary)
Rev1s1on 0 Report Date July 1985 Not Used 2
WCAP-10054-P-A, "Westmghouse Small Break ECCS Evaluation Model usmg the NOTRUMP Code," (W Propnetary)
Rev1s10n 0 Report Date August 1985 Addendum 2, "Addendum to the Westmghouse Small Break ECCS Evaluatron Model Usmg the NOTRUMP Code Safety Injection mto the Broken Loop and COSI Condensatron Model," (W Propnetary) (Referenced m Duke Letter DPC-06-101)
Rev1s10n 1 Report Date 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 Quahficat10n Document for Best-Estimate Loss of Coolant Analysis," (W Propnetary)
Rev1s1on Volume 1 (Rev1s1on 2) and Volumes 2-5 (Rev1s10n 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)
Rev1s10n 1 SER Date January 22, 1991 Rev1s10n 2 SER Dates August 22, 1996 and November 26, 1996 Rev1s1on 3 SER Date June 15, 1994 Not Used
McGmre 1 Cycle 27 Core Operatmg Limits Report 11 Analytical Methods (contmued)
MCEI-0400-379 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, "Multidimens10nal Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary)
Revision 1 Report Date March 2015 8
DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology" Revis10n 4b Report Date September 2010 9
DPC-NE-2004P-A, "Duke Power Company McGmre and Catawba Nuclear Stations Core Thermal-Hydrauhc Methodology usmg VIPRE-01," (DPC Proprietary)
Revision 2a Report Date December 2008 10 DPC-NE-2005P-A, "Thermal Hydrauhc Statistical Core Design Methodology," (DPC Proprietary)
Revis10n 5 Report Date March 2016 11 DPC-NE-2008P-A, "Fuel Mechamcal Reload Analysis Methodology Usmg TAC03," (DPC Propnetary)
Revis10n 0 Report Date April 3, 1995 Not Used 12 DPC-NE-2009-P-A, "Westmghouse Fuel Transit10n 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 1 Cycle 27 Core Operatmg Limits Report 11 Analytical Methods (contmued)
MCEI-0400-379 Page 7 Rev1s10n 0 14 DPC-NF-2010-A, "Duke Power Company McGmre Nuclear Statton Catawba Nuclear Statton Nuclear Physics Methodology for Reload Design "
Rev1s10n 2a Report Date December 2009 15 DPC-NE-2011-PA, "Duke Power Company Nuclear Design Methodology Report for Core Operatmg L1m1ts ofWestmghouse Reactors," (DPC Propnetary)
Rev1s1on la Report Date June 2009 16 DPC-NE-1005-PA, "Nuclear Design Methodology Usmg CASM0-4 / SIMULATE-3 MOX,"
(DPC Propnetary)
Rev1s1on 1 Report Date November 12, 2008 17 DPC-NE-1007-PA, "Cond1t10nal Exemptton of the EOC MTC Measurement Methodology,"
(DPC and W Propnetary)
Rev1s10n 0 Report Date Apnl 2015 18 WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," (W Propnetary)
Rev1s10n 0 Report Date Apnl 1995 19 WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'," (W Propnetary)
Rev1s1on 0 Report Date July 2006
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report 2 0 Operatmg Limits MCEI-0400-379 Page 8 Rev1s1on 0 Cycle-specific parameter hmits for the specifications hsted m Section 1 0 are presented m the followmg subsections These hmits have been developed usmg the NRC approved methodologies specified m Section 1 1 2 1 Reactor Core Safety Limits (TS 2 11) 2 1 1 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 2: 1 3% AK/Km MODE 2 with k-eff < 1 0 and m MODES 3 and4 2 2 2 For TS 3 1 1, SDM shall be 2: 1 0% Af(fK m MODE 5 2 2 3 For TS 3 1 4, SDM shall be 2: 1 3% AK/Km MODES 1 and MODE 2 2 2 4 For TS 3 1 5, SDM shall be 2: 1 3% Af(fK 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% AK/Km MODE 1 and MODE 2 with K-eff2::_
1 0 2 2 6 For TS 3 1 8, SDM shall be 2: 1 3% Af(fK m MODE 2 durmg PHYSICS TESTS
670 McGmre 1 Cycle 27 Core Operatmg L1m1ts Report Figure 1 Reactor Core Safety Limits Four Loops m Operation MCEI-0400-379 Page 9 Rev1s1on 0 DO NOT OPERA TE IN THIS AREA ACCEPTABLE OPERATION 580 L--~~~~..__~~~~-'--~~~~_.__~~~~--'-~~~~-----~~~~~
00 02 04 06 08 1 0 1 2 Fraction of Rated Thermal Power
McGmre 1 Cycle 27 Core Operatmg Limits Report 2 3 Moderator Temperature Coefficient - MTC (TS 3 1 3) 2 3 1 The Moderator Temperature Coefficient (MTC) L1m1ts are MCEI-0400-379 Page 10 Rev1s1on 0 MTC shall be less pos1t1ve than the upper hm1ts shown m Figure 2 BOC, ARO, HZP MTC shall be less positive than O 7E-04 M<IK/°F EOC, ARO, RTP MTC shall be less negative than the -4 3E-04 Af(/K/°F lower MTC hm1t 2 3 2 300 PPM MTC Surveillance L1m1t 1s 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 1s less negative than or equal to the 300 PPM SR 3 1 3 2 Surveillance L1m1t, and all benchmark data contamed m the surveillance procedure 1s 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 L1m1t 1s Measured 60 PPM ARO, equ1hbrmm RTP MTC shall be less negative than or equal to -4 125E-04 M<IK/°F Where BOC = Begmmng of Cycle (burnup correspondmg to the most pos1t1ve MTC)
EOC = End of Cycle ARO = All Rods Out HZP = Hot Zero Power R TP = Rated Thermal Power PPM = Parts per million (Boron) 2 4 Shutdown Bank Insertion Limit (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 Limits (TS 3 1 6) 2 5 1 Control banks shall be w1thm the msert1on, sequence, and overlap hm1ts shown m Figure 3 Specific control bank withdrawal and overlap hm1ts as a funct10n of the fully withdrawn pos1t1on are shown m Table 1
10 09 08
=
Q,)...
07 i:.i
~
Ii-<
Q,) o-.
u~ 06 Q,)
0
~~ 05
~ <l 04 i:::i.. "<:!'
a9
~~ 03
- i......
0._,
~
02
- i..
Q,)
"O 0
01
~
00 McGmre 1 Cycle 27 Core Operatmg Limits Report Figure 2 MCEI-0400-379 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 hm1ts Refer to OP/1/A/6100/22 Umt 1 Data Book for details
231 220 -
200
'c'
- i= 180 c::
-=...
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report Figure 3 MCEI-0400-379 Page 12 Rev1s1on 0 Control Bank Insertion L1m1ts Versus Percent Rated Thermal Power F 11 Wtldr uy 11 a,~~
(Maxtmum=23 l)
~
~
~
~ -
/
/
/
/
/
/
/
Fully Withdrawn
/
.,, : ControlBankB (Mtmmum=222)
I/
- c100% 161) F
/
'= 160 j
Fl (0% 163) 1
/
V
/
- c. 140
/
/
..s
~
=
Q
-e Q
~ =
Q a...
~
C....
-=
Q
~
ControlBank C V
120
/
V V
/
/
100
/
/
/
/
80 V
V I
/
/
ControlBankD I/
/
60
/
/
40 l=1 (0%,47)
V
/
20
.__ :JFully Inserted
/
/
~
I 1(30% 0) I
~
0 -
I I
I I
I/
0 10 20 30 40 50 60 70 80 90 100 Pe1centofRated Thermal Power The Rod Insertion Lmuts (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 < JOO}
Bank CC RIL = 2 3(P) +47
{O < P < 761} for CC RIL = 222 {76 1 < P < JOO}
Bank CB RIL = 2 3(P) + 163 {O < P < 25 7) for CB RIL = 222 {25 7 < P < JOO}
where P = %Rated Thermal Power NOTE Compliance with Techmcal Specification 3 1 3 may reqmre rod withdrawal limits Refer to OP/1/N6100/22 Umt 1 Data Book for details
MCEI-0400-379 Page 13 Rev1s10n 0 McGmre 1 Cycle 27 Core Operatmg Limits 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 Banke BankD Bank A BankB BankC BankD O Start 0
0 0
O 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 0 Start 225 225 116 0 Start 224 224 224 Stop 108 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Fully Withdrawn at 227 Steps Control Control Control Control Control Control Control Control Bank A BankB Banke BankD Bank A BankB BankC BankD O Start 0
0 0
0 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 111 0
226 226 116 0 Start 227 227 116 0 Start 226 226 226 Stop 110 227 227 227 Stop 111 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 O 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 O 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 1 Cycle 27 Core Operatmg L1m1ts Report 2 6 Heat Flux Hot Channel Factor - FQ(X,Y,Z) (TS 3 2 1)
MCEI-0400-379 Page 14 Rev1s1on 0 2 6 1 FQ(X,Y,Z) steady-state limits are defined by the followmg relationships
- where, F ~TP *K(Z)IP F ~TP *K(Z)/0 5 for P > 0 5 for P :SO 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 limits The manufacturmg tolerance and measurement uncertamty are implicitly mcluded m the FQ surveillance limits as defined m Sections 2 6 5 and 2 6 6 2 6 2 F~TP = 2 70 x K(BU) 2 6 3 K(Z) 1s the normalized F Q(X, Y,Z) as a function of core height The K(Z) function for Westmghouse RFA fuel 1s provided m Figure 4 2 6 4 K(BU) 1s the normalized FQ(X,Y,Z) as a function ofburnup F~rPwrth the K(BU) penalty for Westmghouse RF A fuel rs analytically confirmed m cycle-specific reload calculations K(BU) 1s set to I O at all burnups The followmg parameters are reqmred for core momtormg per the Surveillance Reqmrements of Techmcal Specification 3 2 I L
F~(X,Y,Z)
- M0(X,Y,Z) 2 6 5 FQ(X,Y,Z)OP =
UMT :t MT* TILT where Ft (X,Y,Z)OP = Cycle dependent maximum allowable design peakmg factor that ensures F Q(X, Y,Z) LOCA hm1t will be preserved for operation w1thm the AFD, RIL, and QPTR hmrts Ft (X, Y,Z)0P mcludes allowances for calculation and measurement uncertamtres Frf (X,Y,Z) = Design power d1stnbut1on 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 mrtral startup operation
McGmre 1 Cycle 27 Core Operatmg Limits Report MCEI-0400-379 Page 15 Revision 0 MQ(X,Y,Z) = Margm remammg m core location X,Y,Z to the LOCA limit m the transient power distnbut10n MQ(X,Y,Z) is provided m Appendix Table A-1 for normal operatmg conditions and m Appendix Table A-4 for power escalation testmg durmg m1t1al 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 (TIL T = 1 03 5) 2 6 6 F~(X,Y,Z)RPs =
Fg(X,Y,Z)
- Mc(X,Y,Z)
UMT *MT* TILT where L
FQ(X,Y,Z)RPS =
Cycle dependent maximum allowable design peakmg factor that ensures the FQ(X,Y,Z) Centerlme Fuel Melt (CFM) hmrt rs not exceeded for operation wrthm the AFD, RIL, and QPTR L
hmrts FQ(X,Y,Z)RPS mcludes allowances for calculat10n and measurement uncertamtres D
FQ(X,Y,Z) = Defined m Section 2 6 5 Mc(X,Y,Z) = Margm remammg to the CFM hmrt m core location X,Y,Z from the transient power d1stnbut1on Mc(X,Y,Z) rs provided m Appendix Table A-2 for normal operatmg cond1t1ons and m Appendix Table A-5 for power escalat10n testmg durmg m1t1al startup operation UMT = Defined m Section 2 6 5 MT = Defined m Section 2 6 5 TILT = Defined m Section 2 6 5
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report MCEI-0400-379 Page 16 Rev1s1on 0 2 6 7 KSLOPE = 0 0725 where KSLOPE = Adjustment to K 1 value from the OT~ T tnp setpomt required to M
L
~S compensate for each 1 % that FQ (X,Y,Z) exceeds FQ (X,Y,Z) 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
MCEI-0400-379 Page 17 Rev1s10n 0 McGmre 1 Cycle 27 Core Operatmg L1m1ts Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westmghouse RFA Fuel 1 200.--------------------------------,
(0 0, 1 00)
(4 0, 1 00) 1 000... ----------,
0 800
§: 0 600
~
0 400 0 200 0 000 00 Core Height (ft) 00
- '.540
>40 12 0 20 (4 0, 0 9259)
K(Z) 1 0 1 0 0 9259 0 9259 40 60 Core Height (ft)
(12 0, 0 9259) 80 10 0 12 0
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report Table 2 F Q(X, Y,Z) and F MI(X, Y) Penalty Factors MCEI-0400-379 Page 18 Rev1s1on 0 For Techmcal Specification Surveillances 3.2 1 2, 3 2 1 3 and 3 2 2 2 Burnup (EFPD) 4 12 25 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 465 475 499 504 509 519 529 FQ(X,Y,Z)
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 Fm(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 Lmear mterpolat1on 1s adequate for mtermed1ate 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 ~H(X, Y) for compliance with the Techmcal Specification Surveillances 3 2 1 2, 3 2 1 3 and 3 2 2 2
McGmre 1 Cycle 27 Core Operatmg Lnmts Report 2 7 Nuclear Enthalpy Rise Hot Channel Factor - F MJ(X,Y) (TS 3 2 2)
MCEI-0400-379 Page 19 Rev1s1on 0 F Af1 steady-state hm1ts referred to m Techmcal Specification 3 2 2 is defined by the followmg relationship 2 71 FkH(X, Y)LCO= MARP (X,Y) * [ 1 0 + ~
- (1 0 - P)J where FkH (X, Y)Lco is the steady-state, maximum allowed radial peak and mcludes allowances for calculat10n/measurement uncertamty MARP(X,Y) = Cycle-specific operatmg hmit Maximum Allowable Radial Peaks MARP(X, Y) radial peakmg hmits are provided m Table 3 Thermal Power p = Rated Thermal Power RRH = Thermal Power reduction reqmred to compensate for each 1 % the measured radial peak, F~ (X,Y), exceeds its limit (RRH = 3 34 (0 0 < P::: 1 0))
The followmg parameters are reqmred for core momtormg per the surveillance reqmrements ofTechmcal Specification 3 2 2 2 7 2 F~ (X,Y/URV = F~ (X, Y)
- M L\\H (X, Y)
UMR *TILT where F~H (X,Y)SURV --
C I d d
II bl d k
f:
ye e epen ent maximum a owa e esign pea mg actor that ensures the F L\\H(X, Y) limit is not exceeded for operation SURV withm the AFD, RIL, and QPTR hmits F~ (X,Y) mcludes allowances for calculat10n/measurement uncertamty
McGmre 1 Cycle 27 Core Operatmg Limits Report MCEI-0400-379 Page 20 Rev1s1on 0 D
D F 68 (X, Y) = Design radial power distribution for F 68 F 68 (X, Y) is provided m Appendix Table A-3 for normal operation and m Appendix Table A-6 for power escalation testmg durmg mitial startup operat10n M6iX,Y) =Margm remammg m core locat10n X,Y relative to the Operational DNB hmits m the transient power distribution M68 (X,Y) is provided m Appendix Table A-3 for normal operation and m Appendix Table A-6 for power escalation testmg durmg mitial startup operat10n UMR = Uncertamty value for measured radial peaks (UMR = 1 0)
UMR is set to 1 0 smce a factor of 1 04 is implicitly mcluded m the variable Mt1H(X, Y)
TILT = Defined m Sect10n 2 6 5 2 7 3 RRH is defined m Section 2 7 1 2 7 4 TRH= 0 04 where TRH = Reduction m the OT Ll T K1 setpomt reqmred to compensate for each 1 %
the measured radial peak, F~ (X, Y) exceeds its hmit 2 7 5 F Af-1 (X,Y) penalty factors for Techmcal Specification 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) Limits are provided m Figure 5
MCEI-0400-379 Page 21 Rev1s10n 1 McGmre 1 Cycle 27 Core Operatmg Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS)
RFA Steady State Limitmg Value Between Loss of Flow Accident (LOFA) MARPs and FMILOCA Core Axial Peak Height ft 1 05 11 1 2 1 3 14 1 5 1 6 17 1 8 19 2 1 3
3 25 0 12 I 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 3151 1 2461 120 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 I 6058 I 6058 I 6058 I 3007 1 2235 2 40 1 6058 I 6058 I 6058 1 6058 1 6058 1 6058 I 6058 I 6058 I 6058 I 6058 1 6058 1 4633 I 4616 3 60 I 6058 1 6058 I 6058 I 6058 I 6058 I 6058 1 6058 I 6058 I 6058 I 6058 I 6058 1 4675 I 3874 4 80 I 6058 I 6058 1 6058 I 6058 I 6058 1 6058 1 6058 1 6058 1 6058 I 6058 1 6058 I 2987 1 2579 6 00 I 6058 I 6058 I 6058 I 6058 1 6058 1 6058 I 6058 I 6058 1 6058 1 6058 1 6058 I 3293 1 2602 720 I 6058 1 6058 I 6058 I 6058 1 6058 1 6058 I 6058 1 6058 1 6058 1 6058 1 5982 I 2871 1 2195 8 40 I 6058 I 6058 I 6058 I 6058 I 6058 1 6058 1 6058 1 6058 I 6058 1 6010 1 5127 I 2182 1 1578 9 60 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 6058 1 5808 1 5301 1 4444 1 1431 1 0914 10 80 1 6058 I 6058 1 6058 1 6058 1 6058 1 6058 1 5743 1 5573 1 5088 1 4624 1 3832 1 1009 1 0470 1140 1 6058 I 6058 1 6058 I 6058 1 6057 1 5826 1 5289 1 5098 1 4637 1 4218 1 3458 1 0670 1 0142
-50 McGmre 1 Cycle 27 Core Operatmg Limits Report Figure 5 MCEI-0400-379 Page 22 Rev1s1on 0 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits
(-18, 100)
(+10, 100)
Unacceptable Operation 90 Unacceptable Operation 80 Acceptable Operation 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 Specification 3 2 1 may reqmre more restrictive AFD hmits Refer to OP/1/A/6100/22 Umt 1 Data Book for more details
McGmre 1 Cycle 27 Core Operatmg Limits Report MCEI-0400-379 Page 23 Rev1S1on 0 2 9 Reactor Trip System Instrumentation Setpomts (TS 3 3 1) Table 3 3 1-1 2 9 1 Overtemperature AT Setpomt Parameter Values Parameter Nommal Tavg at RTP Nommal RCS Operatmg Pressure Overtemperature AT reactor tnp setpomt Overtemperature AT reactor tnp heatup setpomt penalty coefficient Overtemperature AT reactor tnp depressunzatron setpomt penalty coefficient Time constants utihzed m the lead-lag compensator for AT Time constant utJhzed m the lag compensator for AT Time constants utrhzed m the lead-lag compensator for Tavg Time constant utJhzed m the measured Tavg lag compensator f 1 (Al) "positive" breakpomt f1 (AI) "negatrve" breakpomt f 1 (AI) "positive" slope f1 (AI) "negative" slope T.:::585 l°F P = 2235 psig K1.:S 1 1978 K2 = 0 03341°F K3 = 0 001601/psi
'1 ~ 8 sec
'2.:S 3 sec
,3.:S 2 sec
,4 ~ 28 sec
,5.:S 4 sec
'6.:S 2 sec
= 19 0 %AI
=NIA*
= 1 769 %ATol %AI
=NIA*
f1 (Af) negative breakpomts and slopes for OT~T are less restnct1ve than the OP~T fz(Af) negative breakpomt and slope Therefore, durmg a transient which challenges the negative imbalance hm1ts, OP~T fz(Af) hm1ts will result ma reactor tnp before OT.1T f1 (Af) hm1ts are reached This makes 1mplementat1on of an OT.1T f1 (Af) negative breakpomt and slope unnecessary
MCEI-0400-379 Page 24 Rev1s1on 0 McGmre 1 Cycle 27 Core Operatmg Limits Report 2 9 2 Overpower AT Setpomt Parameter Values Parameter Nommal Tavg atRTP Overpower AT reactor tnp setpomt Overpower AT reactor tnp Penalty Overpower AT reactor tnp heatup setpomt penalty coefficient Time constants utilized m the lead-lag compensator for AT Time constant utilized m the lag compensator for AT Time constant utilized m the measured T avg lag compensator Time constant utilized m the rate-lag controller for T avg fi(AI) "positive" breakpomt fi(AI) "negative" breakpomt fz(AI) "positive" slope fi(.M) "negative" slope T
- 585 1 °P K4:::: 1 0864 K5 = 0 02 I 0 P for mcreasmg Tavg K5 = 0 00 I 0 P for decreasmg Tavg K6 = 0 001179/
0 P for T > T" K6 = 0 0 for T :::: T" t 1 2:. 8 sec t 2 :::: 3 sec t 3 :::: 2 sec
= 35 0 %AI
= -35 0 %AI
= 7 0 %ATof %AI
McGmre 1 Cycle 27 Core Operatmg Limits Report 210 RCS Pressure, Temperature and Flow Limits for DNB (TS 3 41)
MCEI-0400-379 Page 25 Rev1S1on 1 2 10 1 RCS pressure, temperature and flow hmits for DNB are shown m Table 4 211 Accumulators (TS 3 5 1) 2 111 Boron concentration hm1ts durmg MODES 1 and 2, and MODE 3 with RCS pressure > 1000 psi Parameter Applicable Bumup Accumulator mmimum boron concentration 0-200 EFPD Accumulator mmimum boron concentration 200 1 - 250 EFPD Accumulator mmimum boron concentration 250 1 - 300 EFPD Accumulator mmimum boron concentration 300 1 - 350 EFPD Accumulator mmimum boron concentration 350 1 - 400 EFPD Accumulator mmimum boron concentration 400 1 - 450 EFPD Accumulator mimmum boron concentration 450 1 - 465 EFPD Accumulator mmimum boron concentration 465 1 - 519 EFPD Accumulator mmimum boron concentration 519 1 - 529 EFPD Accumulator maximum boron concentration 0- 529 EFPD 212 Refuelmg Water Storage Tank-RWST (TS 3 5 4) 212 1 Boron concentration limits durmg MODES 1, 2, 3, and 4 Parameter R WST mmimum boron concentration RWST maximum boron concentration Limit 2,475 ppm 2,475 ppm 2,468 ppm 2,350 ppm 2,260 ppm 2,188 ppm 2,119 ppm 2,098 ppm 2,014 ppm 2,875 ppm 2,675 ppm 2,875 ppm
1 McGmre 1 Cycle 27 Core Operatmg L1m1ts Report Table 4 Reactor Coolant System DNB Parameters No Operable Parameter lnd1cation 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-0400-379 Page 26 Rev1s10n 0 L1m1ts
~ 587 2 °F
~ 586 9 °F
~ 587 7 °F
~ 587 5 °F
~ 2212 3 ps1g
~2215 0 ps1g
~ 2209 1 ps1g
~ 2211 3 ps1g
~ 390,000 gpm*
- Note The RCS mm1mum coolant flow rate assumed m the hcensmg analyses for the MIC27 core 1s 388,000 gpm However, the flow 1s set at 390,000 gpm, which 1s conservative
McGmre 1 Cycle 27 Core Operatmg L1m1ts Report 2 13 Spent Fuel Pool Boron Concentration (TS 3 7 14)
MCEI-0400-379 Page 27 Rev1s10n 0 2 13 1 Mmimum boron concentration limit for the spent fuel pool Applicable when fuel assemblies are stored m the spent fuel pool Parameter Spent fuel pool mmimum boron concentration 2,675 ppm 2 14 Refuelmg Operations - Boron Concentration (TS 3 9 1) 2 14 1 Mmimum boron concentration limit for the filled portions of the Reactor Coolant System, refuelmg canal, and refuelmg cavity for MODE 6 conditions The mmimum boron concentrat10n limit and plant refuelmg procedures ensure that core Keffremams withm MODE 6 reactivity reqmrement ofKeff :SO 95 Parameter Mmimum boron concentration of the Reactor Coolant System, the refueling canal, and the refuelmg cavity 2,675 ppm
MCEI-0400-379 Page 28 Rev1s10n 0 McGmre 1 Cycle 27 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 (R WST) durmg MODE 4 with any RCS cold leg temperature::: 300 °F and MODES 5 and 6 Parameter Note When cycle burnup 1s > 452 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 RWST mmimum contamed borated water volume RWST mmimum boron concentration RWST mmimum 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-0400-379 Page 29 Rev1s1on 0 McGmre 1 Cycle 27 Core Operatmg L1m1ts Report 2 16 Borated Water Source - Operatmg (SLC 16 9 11) 2 161 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 satisfy UFSAR Chapter 9 reqmrements Parameter Note When cycle burnup 1s > 452 EFPD, Figure 6 may be used to determme reqmred BAT mm1mum level BAT mmimum contamed borated water volume BAT mmimum boron concentration BAT mm1mum water volume reqmred to mamtam SDM at 7,150 ppm RWST mimmum contamed borated water volume R WST mmimum boron concentration RWST maximum boron concentrat10n (TS 3 5 4)
R WST mimmum 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 57,107 gallons 2 17 1 Mm1mum boron concentration hmit for the spent fuel pool reqmred for Standby Makeup Pump Water Supply Applicable for MODES I, 2, and 3 Parameter Spent fuel pool mm1mum boron concentration for TR 16 9 7 2 2,675 ppm
35 o 30 o 25 o MCEI-0400-379 Page 30 Rev1s10n 0 McGmre 1 Cycle 27 Core Operatmg Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus RCS Boron Concentration (Vahd When Cycle Burnup 1s > 452 EFPD)
This figure mcludes addit1onal volumes hsted m SLC 16 9 14 and 16 9 11
{
1 l
J 1
1 1
J I
J I
i i
I RCS Boron II Concentration BAT Level (ppm)
(%level) o_<_3oo I 37 o
-- 300 < 500 -l -~3 0 500 < 700 j__?B 0 700 < 1000 J 23 0
_!Q2.Q_::_130Q_J _1 ~ -~>~13.00 _J ___ 87
't. 20 o
'-" -~~+--~---+-~~:------~-~~---;----~-~-- -- -i----~~--
j I
Acceptable J
1 1
I 1
--r:--! ---:-------i------ : I ll: i ---!~----
l l - :
~[ -!
~
I II J, -
f i
f r
1 J
j r
I I
I 1
j so
-Uo~o=ptar Ope~ooo~,--+~l ___ : 1 ~l ;,,1--t-: --
ll I
I I
I I
~
l I
I I
f l
I I
I 00 +----,-------t--.------,----------.---i-------,----------,---+-------r---i-----,------r---------,
10 o o
200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 RCS Boron Concentration (ppmb)
McGmre 1 Cycle 27 Core Operatmg Limits Report Appendix A Power D1str1but10n Momtormg Factors MCEI-0400-379 Page 31 Rev1s10n 0 Appendix A contams power distribution momtormg factors used m Techmcal Specification Surveillance This data was generated m the McGmre 1 Cycle 27 Maneuvermg Analysis calculation file, MCC-1553 05-00-0664 Due to the size of the momtormg factor data, Appendix A is controlled electromcally withm Duke and is not mcluded m the Duke mtemal copies of the COLR The Plant Nuclear Engmeermg Section will control this mformation via computer file(s) and should be contacted if there is a need to access this mformation Appendix A is mcluded m the COLR copy transmitted to the NRC