ML22306A039

Cycle 23 Core Operating Limits Report

ML22306A039
Person / Time
Site:	South Texas
Issue date:	11/02/2022
From:	Georgeson C South Texas
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NOC-AE-22003924
Download: ML22306A039 (1)

Text

IP.,,..,,

Nuclear Operating Company South Texas Proj ect Electric Gen era ting Sta tion PO Box 2 8 9 Wads worth. Texas 7748]

November 2, 2022 NOC-AE-22003924 10 CFR 50.36 STI: 35384330

ATTN : Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

South Texas Project Unit 2 Docket No. STN 50-499 Unit 2 Cycle 23 Core Operating Limits Report

In accordance with Technical Specification 6.9. 1.6.d, STP Nuclear Operating Company submits

• the attached Core Operating Limits Report (COLR) for Unit 2 Cycle 23. The report covers the core design changes made during the 2RE22 refueling outage.

There are no commitments in this letter.

If there are any questions regarding this report, please contact J. Loya at (361) 972-8218 or me at (361) 972-7806.

JJ!t~

Christopher H. d eorgeson General Manager, Engineering

JAL

Attachment:

South Texas Project Unit 2 Cycle 23 Core Operating Limits Report, Revision 0 NOC-AE-22003924 Page 2 of 2

cc:

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 E. Lamar Boulevard Arlington, TX 76011-4511

Gregory Kolcum Senior Resident Inspector, South Texas Project U.S. Nuclear Regulatory Commission

Chad Scott Resident Inspector, South Texas Project U.S. Nuclear Regulatory Commission

Dennis Galvin Project Manager U.S Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Licensing Project Branch 4 NOC-AE-22003924 Attachment

Attachment

South Texas Project Unit 2 Cycle 23 Core Operating Limits Report, Revision 0 STI :35388293

Nuclear Operating Company ~... --

SOUTH TEXAS PROJECT

Unit 2 Cycle 23

CORE OPERATING LIMITS REPORT

U2C23-0

Revision 0

Core Operatin g Limit s Report Pag e 1 of 17

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 2 of 17

1.0 CORE OPERA TING LIMITS REPORT

This Core Operating Limits Report for STPEGS Unit 2 Cycle 23 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.6.

The Technical Specifications affected by this report are :

1) 2.1 SAFETY LIMITS

2) 2.2 LIMITING SAFETY SYSTEM SETTINGS

3) 3/4.1.1.1 SHUTDOWN MARGIN

4) 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT LIMITS

5) 3/4.1.3.5 SHUTDOWN ROD INSERTION LIMITS

6) 3/4.1.3.6 CONTROL ROD INSERTION LIMITS

7) 3/4.2.1 AFD LIMITS

8) 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR

9) 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

10) 3/4.2.5 DNB PARAMETERS

2.0 OPERATING LIMITS

The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented below.

2.1 SAFETY LIMITS (Specification 2.1):

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 1.

2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2):

2.2.1 The Loop design flow for Reactor Coolant Flow-Low is 98,000 gpm.

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 3 of 17

2.2.2 The Over-temperature i::1T and Over-power i::1T setpoint parameter values are listed below :

Over-temperature L'.1 T Setpoint Parameter Values 11 measured reactor vessel i::1T lead /lag time constant, 11 = 8 sec 12 measured reactor vessel L'.1 T lead /lag time constant, 12 = 3 sec 13 measured reactor vessel L'.1 T lag time constant, 13 = 2 sec 14 measured reactor vessel average temperature lead /lag time constant, 14 = 28 sec 1s measured reactor vessel average temperature lead /lag time constant, 1s = 4 sec 1°6 measured reactor vessel average temperature lag time constant, 1°6 = 2 sec K1 Overtemperature L'.1 T reactor trip setpoint, K1 = 1.14 K2 Overtemperature i::1T reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K3 Overtemperature i::1T reactor trip setpoint pressure coefficient, K3 = 0.00143 /psi T' Nominal full power Tavg, T' :::; 592.0 °F P' Nominal RCS pressure, P' = 2235 psig f1(M) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers ; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For q1 - qb between -70% and +8%, fi(~I) = 0, where q1 and qb are percent RA TED THERMAL POWER in the top and bottom halves of the core respectively, and q1 + qb is total THERMAL POWER in percent of RATED THERMAL POWER ;

(2) For each percent that q1 - qb is more negative than -70%, the ~ T Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THERMAL POWER ;

and (3) For each percent that q1 - qb is more positive than +8%, the~ T Trip Setpoint shall be automatically reduced by 2.65% of its value at RATED THERMAL POWER.

(Reference 3.6 and Section 4.4.1.2 of Reference 3.7)

Over-power i::1T Setpoint Parameter Values 11 measured reactor vessel i::1T lead /lag time constant, 11 = 8 sec 12 measured reactor vessel L'.1 T lead /lag time constant, 12 = 3 sec 13 measured reactor vessel L'.1 T lag time constant, 13 = 2 sec 1°6 measured reactor vessel average temperature lag time constant, 1°6 = 2 sec 1 1 Time constant utilized in the rate-lag compensator for Tavg, 11 = 10 sec K4 Overpower L'.1 T reactor trip setpoint, K4 = 1. 08 Ks Overpower L'.1 T reactor trip setpoint T avg rate /lag coefficient,

Ks= 0.02/°F for increasing average temperature, and Ks = 0 for decreasing average temperature K6 Overpower L'.1 T reactor trip setpoint T avg heatup coefficient K6 = 0.002/°F for T > T", and K6 = 0 for T :::; T" T" Indicated full power Tavg, T":::; 592.0 °F fi(M) = 0 for all (M)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 4 of 17

2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):

The SHUTDOWN MARGIN shall be :

2.3.1 Greater than 1.3% L'.1p for MODES 1 and 2*

• See Special Test Exception 3.10.1 2.3.2 Greater than the limits in Figure 2 for MODES 3 and 4.

2.3.3 Greater than the limits in Figure 3 for MODE 5.

2.4 MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3):

2.4.1 The BOL, ARO, MTC shall be less positive than the limits shown in Figure 4.

2.4.2 The EOL, ARO, HFP, MTC shall be less negative than-62.6 pcm/°F.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than-53.6 pcm/°F (300 ppm Surveillance Limit).

Where : BOL stands for Beginning-of-Cycle Life,

EOL stands for End-of-Cycle Life,

ARO stands for All Rods Out,

HFP stands for Hot Full Power (100% RATED THERMAL POWER),

HFP vessel average temperature is 592 °F.

2.4.4 The Revised Predicted near-EOL 300 ppm MTC shall be calculated using the algorithm from the document referenced by Technical Specification 6.9.1.6.b.10 :

Revised Predicted MTC = Predicted MTC + AFD Correction - 3 pcm/°F

If the Revised Predicted MTC is less negative than the COLR Section 2.4.3 limit and all of the benchmark data contained in the surveillance procedure are met, then an MTC measurement in accordance with S.R. 4.1.1.3b is not required.

2.5 ROD INSERTION LIMITS (Specification 3.1.3.5 and 3.1.3.6):

2.5.1 All banks shall have the same Full Out Position (FOP) of either 257 or 259 steps withdrawn.

2.5.2 The Control Banks shall be limited in physical insertion as specified in Figure 5.

2.5.3 Individual Shutdown bank rods are fully withdrawn when the Bank Demand Indication is at the FOP and the Rod Group Height Limiting Condition for Operation is satisfied (T.S. 3.1.3.1).

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 5 of 17

2.6 AXIAL FLUX DIFFERENCE (Specification 3.2.1):

2.6.1 AFD limits as required by Technical Specification 3.2.1 are determined by Constant Axial Offset Control (CAOC) Operations with an AFD target band of +5, -10%.

2.6.2 The AFD shall be maintained within the ACCEPTABLE OPERATION portion of Figure 6, as required by Technical Specifications.

2.7 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):

2.7.1 F~TP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (F~J P) within specific core planes shall be :

2.7.3.1 Less than or equal to 2.102 for all cycle bumups for all core planes containing Bank "D" control rods, and 2.7.3.2 Less than or equal to the appropriate core height-dependent value from Table 1 for all unrodded core planes.

2.7.3.3 PF xy = 0.2.

These F xy limits were used to confirm that the heat flux hot channel factor FQ(Z) will be limited by Technical Specification 3.2.2 assuming the most-limiting axial power distributions expected to result for the insertion and removal of Control Banks C and D during operation, including the accompanying variations in the axial xenon and power distributions, as described in WCAP-8385. Therefore, these Fxy limits provide assurance that the initial conditions assumed in the LOCA analysis are met, along with the ECCS acceptance criteria of 10 CFR 50.46.

2.7.4 The Fxy limits from Section 2.7.3 above are not applicable in the following core plane regions :

2.7.4.1 Upper and lower core plane regions as presented in Table 2, and 2.7.4.2 Grid plane regions as presented in Table 2, and 2.7.4.3 Core plane regions within+/- 2% of core height(+/- 3.36 inches) about the bank demand position of the bank " D " control rods.

2.7.5 Core Power Distribution Measurement Uncertainty for the Heat Flux Hot Channel Factor 2.7.5.1 If the Power Distribution Monitoring System (PDMS) is operable, as defined in the Technical Requirements Manual Section 3.3.3.12, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and Fxy (Z) using the PDMS shall be calculated by :

UFQ = (1.0 + (UQ/lO0))*UE Where :

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 6 of 17

UQ = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced by Technical Specification 6.9.1.6.b.11 V E = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System (PDMS).

2.7.5.2 If the moveable detector system is used, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and Fxy (Z) shall be calculated by :

UFQ=UQu*UE Where :

UQu = Base FQ measurement uncertainty of 1.05.

V E = Engineering uncertainty factor of 1.03.

2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):

2.8.1 FfJP = 1.62 2.8.2 PF t,H = 0.3 2.8.3 Core Power Distribution Measurement Uncertainty for the Enthalpy Rise Hot Channel Factor 2.8.3.1 If the Power Distribution Monitoring System (PDMS) is operable, as defined in the Technical Requirements Manual Section 3.3.3.12, the core power distribution measurement uncertainty (UMH) to be applied to the F~H using the PDMS shall be the greater of:

UFMI = 1.04 OR UFMI = 1.0 + (Um/100)

Where :

u,,,H = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced in Technical Specification 6.9.1.6.b.11.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System.

2.8.3.2 If the moveable detector system is used, the core power distribution measurement uncertainty (UMH) shall be :

UFm= 1.04

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 7 of 17

2.9 DNB PARAMETERS (Specification 3.2.5):

2.9.1 The following DNB-related parameters shall be maintained within the following limits (Nominal Values from Reference 3.1, as annotated below) : 1 2.9.1.1 Reactor Coolant System Tav g :::; 595 °F 2,

2.9.1.2 Pressurizer Pressure > 2200 psig 3,

2.9.1.3 Minimum Measured Reactor Coolant System Flow > 403,000 gpm 4.

3.0 REFERENCES

3.1 Letter from A. S. Ganey (Westinghouse) to F. Yilmaz (STPNOC), "South Texas Project Electric Generating Station Unit 2 Cycle 23 Final Reload Evaluation" NF-TG-22-028 (ST-UB-NOC-22000019) dated August 1, 2022. (STI 35385597) 3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. 1 and 2.

3.3 STPNOC Calculation ZC07035, Rev. 3, "Loop Uncertainty Calculation for RCS Tavg Instrumentation," Section 10.1. (STI 35343289) 3.4 STPNOC Calculation ZC07032, Rev. 7, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9. (STI 35191115) 3.5 Letter from J. S Wyble (Westinghouse) to T. J. Jordan (STPNOC), "STP Nuclear Operating Company Units 1 & 2 Power Uprate PCWG Parameters," ST-WN-NOC-00-000072 dated December 15, 2000. (STI 31218644) 3.6 Letter from J.M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Units 1 and 2 Documentation of the f1(~I) Function in OT~T Setpoint Calculation," NF-TG-11-93 (ST-UB-NOC-11003215) dated November 10, 2011.

(STI 33079478) 3.7 Document RSE-U2, Rev. 11, "Unit 2 Cycle 23 Reload Safety Evaluation and Core Operating Limits Report." (CR Action 21-4948-55)

A discussion of the processes to be used to take these readings is provided in the basis for Technical Specification 3.2.5.

2 Includes a 1.9 °F measurement uncertainty per Reference 3.3, Page 3 7.

Limit not applicable during either a Thermal Power ramp in excess of 5% of RTP per minute or a Thermal Power step in excess of 10% RTP. Per Technical Specification 3.2.5 Bases, this includes a 10.7 psi measurement uncertainty as read on the QDPS display, which when added to the safety analysis limit of 2189 psig gives the COLR limit listed above.

The 10.7 psi uncertainty is bounded by the 9.6 psi averaged measurement calculated in Reference 3.4.

4 Includes the flow measurement uncertainty of2.8% from Reference 3.5.

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 8 of 17

Figure 1

Reactor Core Safety Limits - Four Loops in Operation

680

i c 2,664.52 ) I

-..... r--,... I Unacceptable 660............,_

I c 2, 652. 10 ) I"'---,........

.................... r--,... 1........ 2450 PSIAI I"'---, -- ---I"'---,

~........ -

-..................... r--,...........

640........ --- C 96, 638.57 ) I I"'---, 12 250 PSIAI "

- -........... \\. r---........... "'~

I c 2, 622 _54 )........ --- --- "

620..... ""'-- --., I"'-.... "' \\.

~ I ( 102, 622.62 ) I I\\. '\\.

e_., ~ -......

~........ "', "\\"'-~ -.....

E-< cs:........ 1830 PSIAI '\\. ' r---... ~I"'-....

r/1 c 130, 600. 11 ) I

~ 600........ "\\........ "'\\

........ _ \\.

I"""'--,............ \\. c 130, 589.08 ) I

i"""'-,. "'\\.

I ( 112, 587.35 ) I "' \\.

580 ~\\.

\\

\\. c 130, 563.93 ) I

! Acceptable !

560

540 0 20 40 60 80 100 120 140 Rated Thermal Powe r(%)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 9 of 17

Figure 2

Required Shutdown Margin for Modes 3 & 4

7.0

6. 0

I I I Acceptable I c 2400, s. 1 s )I

5. 0 /

,, /"

V /

/

~ V 0 ~v

~V ~v

/ /

0 /

I/ V

/ /

2. 0 /r'

/~

/"

1co, uo )1 / /,, Unacceptable !

IC 600, 1.30 ) I

1. 0

0. 0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 10 of 17

Figure 3

Required Shutdown Margin for Mode 5

7.0

6.0

I I I Acceptable I

5.0

1( 2400 ; 4.50 ) I

./

/,,

/,,

/"'

/

/ /

,.V V

~ V

,, ~ /

./

2.0 /.,,, /,,

/

l( O ; 1.30 ) 1 / ! Unacceptable !

~

1.0 IC 6so ; uo )

0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 11 of 17

Figure 4

MTC versus Power Level

7.0

6.0 I unacceptable

I co, 5.o) I oo, 5.o) I 5.0

\\ \\

\\ \\

I I,

I Acceptable I \\

\\ \\

' " \\

' \\

\\

looo, o.o)I

-1.0

-2.0

-3.0 0 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 12 of 17

Figure 5

Control Rod Insertion Limits* versus Power Level 260,,,, :c 23, 259): 122 Step Overlap I, I c "'I 122 Step Overlap I 79, 259 ) :

,, I ( 22, 257 ): 120 Step Overlap, I ( 78, 257 ):,. 120 Step Overlap

1,

240,,,,,

I 220,, 1BankB "'

200.... ( o, 202 )I

180,J "" ( 100, 174) ~ ~

,J' / ""

160,J / " I/

-"',J BankC,I Q.,,J /

Q) oo 140 l,J I/

'-" l,J /

=,J I/

..... 0 I/ /

-"',J /

~ 0 120 l,J /

Q., I/ V

=

0 / /

~ 100 - I/ I/

"C / /

0,I /

~,I V 1BankD1 80 I/,,

I/ /,

/, V 60 ( 0, 65),,,,

40,.,,

,.,

• Control Bank A is already withdrawn to Full Out Position.,.,

20,.,, Fully withdrawn shall be the condition where shutdown and

( 29, 0),, "' control banks are at the position of either 257 or 259 steps 0 "" withdrawn. I I I I

0 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 13 of 17

Figure 6

AFD Limits versus Power Level

120

110

100

( -11, 90) I I I ( 11, 90) I 90, ~

, ' ' \\

80, ' ' "

-~, ' ~ - ~

j Unacceptable ' " i Unacceptable

~ \\ '

-~ r I I

-=.. 70, : Acceptable: "

~

~ ', '

0 "

~

-; 60, ' ' "

s ' \\ '

~ '

-= \\

E-, ', ~ "

"C

..... J ~ 50

~

~ (-31, 50) : ( 31, 50):

40

30

20

10

0

-50 -40 20 -10 0 10 20 30 40 50

Axial Flux Difference (% Delta-I)

Unit 2 Cycle 23 Nuclear Operating Company Core Operating Limits Report Rev. 0

... **-.-- U2C23-0 Page 14 of 17

Figure 7

K(Z) - Normalized FQ(Z) versus Core Height

1.2 -* I i ** \\ I

i I

- I i 1.1 - I. - I !

- I I I : -

I I : I i I 1.0 I I i I I

' I - ;

I I i r ' I 0.9 I ' j '

-* I ' I ' I i ' I I i 0.8 -

§: I,. J I

~.. ~ 0.7 I I I I ' I I I

-+-> 0 I I CJ ' I -

~

~..

OJ) 0.6 I I I I

= -* i l

~ - Core Elev. (ft) F Q K(Z) : 1,__

~

~ I 0.0 2.55 1.0 __

~ --- ',_

"t:I 0.5 7.0 2.55 1.0 I c.) I -.L N 14.0 2.3 58 0.925 1 =-= I ~ I s i

.. 0.4 ' I 0 I _... I I ' I ---

z

i 0.3 i *I* ! i I

I i

! 1 i 0.2 I I i I I I I

I - 1-1 i

\\ I 0.1 I !

• - ---r

- i ! I I --

1_

- I ' - --

0.0 0 2 3 4 5 6 7 8 9 11 12 13 14 Core Height (ft)

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 15 of 17

Table 1 (Part 1 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Less Than 9,000 MWD/MTU (Subject to Exclusion Zones ofCOLR Section 2.7.4)

Core Height Axial Unrodded Core Height Axial Unrodded Ft. Point Fx Ft. Point Fx 14.0 1 7.514 6.8 37 1.954 13.8 2 5.846 6.6 38 2.000 13.6 3 4.182 6.4 39 1.977 13.4 4 2.763 6.2 40 1.930 13.2 5 2.551 6.0 41 1.891 13.0 6 2.247 5.8 42 1.906 12.8 7 2.132 5.6 43 1.918 12.6 8 2.113 5.4 44 1.920 12.4 9 2.063 5.2 45 1.961 12.2 10 2.002 5.0 46 2.028 12.0 11 1.965 4.8 47 2.050 11.8 12 1.976 4.6 48 1.998 11.6 13 2.020 4.4 49 1.934 11.4 14 2.004 4.2 50 1.947 11.2 15 1.948 4.0 51 1.953 11.0 16 1.912 3.8 52 1.945 10.8 17 1.904 3.6 53 1.956 10.6 18 1.894 3.4 54 2.005 10.4 19 1.883 3.2 55 2.042 10.2 20 1.906 3.0 56 1.993 10.0 21 1.957 2.8 57 1.945 9.8 22 1.977 2.6 58 1.952 9.6 23 1.928 2.4 59 1.957 9.4 24 1.877 2.2 60 1.966 9.2 25 1.874 2.0 61 2.004 9.0 26 1.865 1.8 62 2.086 8.8 27 1.861 1.6 63 2.157 8.6 28 1.873 1.4 64 2.164 8.4 29 1.933 1.2 65 2.197 8.2 30 1.988 1.0 66 2.322 8.0 31 1.959 0.8 67 2.741 7.8 32 1.920 0.6 68 3.719 7.6 33 1.916 0.4 69 5.352 7.4 34 1.922 0.2 70 7.633 7.2 35 1.921 0.0 71 11.405 7.0 36 1.915

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 16 of 17

Table 1 (Part 2 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 9,000 MWD/MTU (Subject to Exclusion Zones ofCOLR Section 2.7.4)

Core Height Axial Unrodded Core Height Axial Unrodded Ft. Point Fx Ft. Point Fx 14.0 1 6.583 6.8 37 2.203 13.8 2 5.240 6.6 38 2.245 13.6 3 3.898 6.4 39 2.208 13.4 4 2.751 6.2 40 2.144 13.2 5 2.598 6.0 41 2.108 13.0 6 2.314 5.8 42 2.100 12.8 7 2. 155 5.6 43 2.088 12.6 8 2.089 5.4 44 2.072 12.4 9 2.025 5.2 45 2.092 12.2 10 2.012 5.0 46 2.132 12.0 11 2.017 4.8 47 2.130 11.8 12 2.052 4.6 48 2.069 11.6 13 2. 100 4.4 49 2.020 11.4 14 2.087 4.2 50 2.015 11.2 15 2.044 4.0 51 2.003 11.0 16 1.995 3.8 52 1.988 10.8 17 2.018 3.6 53 1.990 10.6 18 2.032 3.4 54 2.029 10.4 19 2.033 3.2 55 2.057 10.2 20 2.066 3.0 56 1.999 10.0 21 2. 126 2.8 57 1.946 9.8 22 2. 153 2.6 58 1.926 9.6 23 2. 111 2.4 59 1.902 9.4 24 2.076 2.2 60 1.886 9.2 25 2.091 2.0 61 1.896 9.0 26 2. 102 1.8 62 1.940 8.8 27 2. 108 1.6 63 1.970 8.6 28 2. 124 1.4 64 1.947 8.4 29 2. 177 1.2 65 1.960 8.2 30 2.227 1.0 66 2.071 8.0 31 2. 183 0.8 67 2.443 7.8 32 2. 142 0.6 68 3.187 7.6 33 2. 143 0.4 69 4.320 7.4 34 2. 154 0.2 70 5.873 7.2 35 2. 161 0.0 71 8.603 7.0 36 2. 162

.,.. Unit 2 Cycle 23 Nuclear Operat ing Company Core Operating Limits Report Rev. 0

... ~ U2C23-0 Page 17 of 17

Table 2 Core and Grid Plane Fxy Exclusion Zones Core Core Core Core Axial Height Height Axial Height Height Point (in.) (%) Top /Bottom Grid Point (in.) (%) Top /Bottom Grid 1 168.0 100.0 Excluded 37 81.6 48.6 2 165.6 98.6 Excluded 38 79.2 47.1 Excluded 3 163.2 97. 1 Excluded 39 76. 8 45.7 Excluded 4 160.8 95.7 Excluded 40 74.4 44.3 5 158.4 94.3 Excluded Excluded 41 72.0 42.9 6 156.0 92.9 Excluded Excluded 42 69.6 41.4 7 153.6 91.4 Excluded 43 67.2 40.0 8 151.2 90.0 Excluded 44 64. 8 38.6 9 148.8 88.6 45 62.4 37.1 10 146.4 87. 1 46 60.0 35.7 Excluded 11 144.0 85.7 47 57.6 34.3 Excluded 12 141.6 84.3 48 55.2 32.9 Excluded 13 139.2 82.9 Excluded 49 52. 8 31.4 14 136.8 81.4 Excluded 50 50.4 30.0 15 134.4 80.0 Excluded 51 48.0 28.6 16 132.0 78.6 52 45.6 27.1 17 129.6 77. 1 53 43.2 25.7 18 127.2 75.7 54 40. 8 24.3 Excluded 19 124.8 74.3 55 38.4 22.9 Excluded 20 122.4 72.9 56 36.0 21.4 Excluded 21 120.0 71.4 Excluded 57 33.6 20.0 22 117.6 70.0 Excluded 58 31.2 18.6 23 115.2 68.6 Excluded 59 28. 8 17.1 24 112.8 67. 1 60 26.4 15.7 25 110.4 65.7 61 24.0 14.3 26 108.0 64.3 62 21.6 12.9 Excluded 27 105.6 62.9 63 19.2 11.4 Excluded 28 103.2 61.4 64 16. 8 10.0 Excluded Excluded 29 100.8 60.0 Excluded 65 14.4 8.6 Excluded 30 98.4 58.6 Excluded 66 12.0 7.1 Excluded 31 96.0 57. 1 Excluded 67 9.6 5.7 Excluded 32 93.6 55.7 68 7.2 4.3 Excluded 33 91.2 54.3 69 4. 8 2.9 Excluded Excluded 34 88. 8 52.9 70 2.4 1.4 Excluded Excluded 35 86.4 51.4 71 0.0 0.0 Excluded Excluded 36 84.0 50.0