NOC-AE-09002486, Cycle 16 Core Operating Limits Report

Cycle 16 Core Operating Limits Report

ML093210176
Person / Time
Site:	South Texas
Issue date:	11/12/2009
From:	Leazar D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-09002486
Download: ML093210176 (25)

Text

Nuclear Operating Company South Texas ProjectElectric GeneratingStation PO Boxr 289 Wadsvorth. Texas 77483 -

November 12, 2009 NOC-AE-09002486 10CFR50.36 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 South Texas Project Unit 1 Docket No. STN 50-498 Unit 1 Cycle 16 Core Operating Limits Report In accordance with Technical Specification 6.9.1.6.d, STP Nuclear Operating Company submits the attached Revision 0 and Revision 1 to the Core Operating Limits Report for Unit 1 Cycle 16. The reports reflect core design changes made during the 1RE15 refueling outage.

There are no commitments in this letter.

If there are any questions concerning this report, please contact Robyn Savage at (361) 972-7438 or me at (361) 972-7795.

David A. LeazaI/

Manager, Nuclear Fuels & Analysis rds

Attachment:

Unit 1 Cycle 16 Core Operating Limits Report, Rev. 0 Unit 1 Cycle 16 Core Operating Limits Report, Rev. 1 42w ST1 32572965

NOC-AE-09002486 Page 2 of 2 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 612 East Lamar Blvd, Suite 400 Arlington, Texas 76011-8064 Mohan C. Thadani U. S. Nuclear Regulatory Commission Mohan C. Thadani Kevin Howell Senior Project Manager Catherine Callaway U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1A) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector Ed Alarcon U. S. Nuclear Regulatory Commission J. J. Nesrsta P, O. Box 289, Mail Code: MN116 R. K. Temple Wadsworth, TX 77483 Kevin Polio City Public Service C. M. Canady Jon C. Wood City of Austin Cox Smith Matthews Electric Utility Department 721 Barton Springs Road C. Mele Austin, TX 78704 City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services

Ti M" Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 0 W A " Page 1 of 7 Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 16 CORE OPERATING LIMITS REPORT Revision 0 MODES 3. 4 and 5

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 0 W EA r Page 2 of 7 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit I Cycle 16 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 Will be added prior to Mode 2.

2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2):

2.2.1 'Will be added prior to Mode 2.

2.2.2 The Over-temperature AT and Over-power AT setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values Will be added prior to Mode 2.

Over-power AT Setpoint Parameter Values Will be added prior to Mode 2.

2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):

The SHUTDOWN MARGIN shall be:

2.3.1 Will be added prior to Mode 2.

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.

OwqAM Nuclear Operating Company Unit 1 Cycle 16 Core Operating Limits Report Rev. 0 E N Page 3of 7 2.4 MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3):

2.4.1 Will be added prior to Mode 2.

2.4.2 Will be added prior to Mode 2.

2.4.3 Will be added prior to Mode 2.

2.4.4 Will be added prior to Mode 2.

2.5 ROD INSERTION LIMITS (Specification 3.1.3.5 and 3.1.3.6):

2.5.1 Will be added prior to Mode 2.

2.5.2 Will be added prior to Mode 2.

2.5.3 Will be added prior to Mode 2.

2.6 AXIAL FLUX DIFFERENCE (Specification 3.2.1):

2.6.1 Will be added prior to Mode 2.

2.6.2 Will be added prior to Mode 2.

2.7 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):

2.7.1 Will be added prior to Mode 2.

2.7.2 Will be added prior to Mode 2.

2.7.3 Will be added prior to Mode 2.

2.7.4 Will be added prior to Mode 2.

2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.23):

2.8.1 Will be added prior to Mode 2.

2.8.2 Will be added prior to Mode 2.

2.8.3, Will be added prior to Mode 2.

2.9 DNB PARAMETERS (Specification 3.2.5):

2.9.1 Will be added prior to Mode 2.

3.0 REFERENCES

3.1 Letter from D. V. Lockridge (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 1 Cycle 16 Final Reload Evaluation (RE)," NF-TG-09-74 (ST-UB-NOC-09003012), August 20, 2009.

3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. 1 and 2.

3.3 CREE 08-8174-79, Reload Safety Evaluation and COLR, Unit 1 Cycle 16, Modes 3, 4 and 5.

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 0 E B Page4 of 7 FIGURE 1 Reactor Core Safety Limits - Four Loops in Operation Will be added prior to Mode 2.

Unit 1 Cycle 16 Operating Company Core Operating Limits Report Rev. 0

IF Mr - Page 5 of 7 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0 S4.0 a..*

' 3.0 2.0 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 0 W E N-- , Page 6 of 7 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 S4.0

.Ei. 3.0 2.0 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 0 iN ur Page 7 of 7 FIGURE 4 MTC versus Power Level Will be added prior to Mode 2.

FIGURE 5 Control Rod Insertion Limits versus Power Level Will be added prior to Mode 2.

FIGURE 6 AFD Limits versus Power Level Will be added prior to Mode 2.

FIGURE 7 K(Z) - Normalized FQ(Z) versus Core Height Will be added prior to Mode 2.

TABLE 1 Unrodded Fxy for Each Core Height Will be added prior to Mode 2.

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 16 CORE OPERATING LIMITS REPORT Revision I Core Operating Limits Report Page I of 16

• llll!*Unitr1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 16 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.

WTM Nuclear Operating company Unit 1Cycle 16 Core Operating Limits Report Rev. 1 Page 3 of 16 2.2.2 The Over-temperature AT and Over-power AT setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values t1 measured reactor vessel AT lead/lag time constant, "T,= 8 sec T2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, T4 = 28 sec "T5 measured reactor vessel average temperature lead/lag time constant, T5 = 4 sec T6 measured reactor vessel average temperature lag time constant, "r6 = 2 sec K, Overtemperature AT reactor trip setpoint, K, = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.00143/psig T' Nominal full power Tavg, T'< 592.0 'F P' Nominal RCS pressure, P' = 2235 psig f, (AI) is a function of the indicated difference between top and bottom detectors of the r power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For qt - qb between -70% and +8%, fl(AI) = 0, where q% and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qb exceeds -70%, the AT Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of q, - qb exceeds +8%, the AT Trip Setpoint shall be automatically reduced by 2.65% of its value at RATED THERMAL POWER.

Over-power AT Setpoint Parameter Values "r measured reactor vessel AT lead/lag time constant, T1 = 8 sec.

-r2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, x3 = 2 sec T6 measured reactor vessel average temperature lag time constant, t 6 = 2 sec

'17 Time constant utilized in the rate-lag compensator for Tavg, t 7 = 10 sec K4 Overpower AT reactor trip setpoint, K4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, K5 = 0.02/°F for increasing average temperature, and K5 = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient K(6 = 0.002/°F for T>T",and K6 = 0forT_< T" T" Indicated full power Tavg, T"< 592.0 'F f 2(AI) = 0 for all (Al)

TiCm Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 4 of 16 2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):

The SHUTDOWN MARGIN shall be:

2.3.1 Greater than 1.3% Ap 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 T.S. 6.9.1.6.b. 10:

Revised Predicted MTC = Predicted MTC + AFD Correction - 3 pcm/IF If the Revised Predicted MTC is less negative than the S.R. 4.1.1.3b 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 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).

T M Nuclear Operating Company Unit 1Cycle 16 Core Operating Limits Report Rev. 1

- rPage 5 of 16 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):

RTP 2.7.1 FQ -2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (F*Ty) within specific core planes shall be:

2.7.3.1 Less than or equal to 2.102 for all cycle burnups 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 PFY = 0.2.

These Fxy 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 F.y 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 Core Power Distribution Measurement Uncertainty for the Heat Flux Hot Channel Factor 2.7.4.1 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/IOO))*UE Where:

UQ = Uncertainty for power peaking factor as defined in Equation 5-19 of Reference 3.6 UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the BEACON computer code.

Unit 1 Cycle 16 Nuclear perating Company Core Operating Limits Report Rev. 1 Page 6 of 16 2.7.4.2 If the moveable detector system is used, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and F,(Z) shall be calculated by:

UFQ = UQU*UE Where:

UQU Base FQ measurement uncertainty of 1.05.

UE = Engineering uncertainty factor of 1.03.

2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):

FRTP1 2.8.1 FAH= 1.62' 2.8.2 PFAH 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 (UFAH) to be applied to the FNH using the PDMS shall be calculated by:

UFAH = 1.0 + (UAH/100)

Where:

UAH= Uncertainty for power peaking factor as defined in Equation 5-19 of Reference 3.6 This uncertainty is calculated and applied automatically'by the BEACON computer code.

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

UFAH = 1.04 Applies to all fuel in the Unit 1 Cycle 16 Core.

• lDT~l*Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev.1 Page 7 of 16 2.9 DNB PARAMETERS (Specification 3.2.5):

2.9.1 The following DNB-related parameters shall be maintained within the following limits:

2.9.1.1 Reactor Coolant System Tavg < 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 D. V. Lockridge (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 1 Cycle 16 Final Reload Evaluation (RE)," NF-TG-09-74 (ST-UB-NOC-09003012), August 20, 2009.

3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. 1 and 2.

3.3 STPNOC Calculation ZC-7035, Rev. 2, "Loop Uncertainty Calculation for RCS Tavg Instrumentation," Section 10.1, effective July 22, 2003.

3.4 STPNOC Calculation ZC-7032, Rev. 4, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9, effective July 22, 2003.

3.5 Condition Report Engineering Evaluation 08-8174-9, Revision 0, "Reload Safety Evaluation and Core Operating Limits Report for South Texas Unit 1-Cycle 16 Modes 1, 2, 3, 4, and 5."

3.6 WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

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.

3 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. Includes a 9.6 psi measurement uncertainty as read on QDPS display per Reference 3.4.

4 Includes a 2.8% flow measurement uncertainty.

I TM Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 8 of 16 Figur 1 Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 i-600 580 560 540 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

ATI MAN Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0

~4.0

~3.0 2.0 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

MMf Unit 1 Cycle 16 IN!clear Opefrating Company Core Operating Limits Report Rev. 1 soINN-- Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 Z

S4.0 3.0 2.0 F(C 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

AT M Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 SUnacceptable Operation 5.0 4.0 I Acceptable Operation S3.0 T\

I--

U o 2.0

0. __ __ _ _ _t_ _ _ _ _ _ _ _

a-a-

C o 00

-1.0

-2.0

-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

lil Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 SO AVM" Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 2322959 12 Sep1

, ItL 1 79,259 122 Step Overrlap 240

.. B~ank O . . cBI1f I I 1r1 1 11 1 11 1 1z Zi hi

-I__

220 i ~i ' .iJ I.. I I[I 200 180

' 160 I - I L 202 140 IýLI

= 120 ____J Ban CBn

__I Ii!z I zl it] -

100 80 I OF F wll ith t ls b t c w 0'l 00 Coto akAis aled ithrwJ oF l u oiin 60 lywthrw shallb h odto hresudw n contro bakr Control saraywt ~Bank A ttepstono29seswtdan DrwtoFlOtPsion b"II I a a I 40 20 0

0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

A T M Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110 L(11,91)

II 100 90 Ill I~ 111 - - - - -

80 Unacceptable \1 Unacceptable 70 Operation (-1 ,90 Acceptable I I. Operation

. .3..

a- I 77 I C

2* 60 1'1 LH t IIII 50 jl--I I-------------- I i i]

1# r 40 30 20 10 0

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Flux Difference (% Delta-I)

• IN~p~rin CUnit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 1.1 i : i i: , i i i I i i i i i ; i , i I I I I I II I I I i1 1.0 i  ! I  ! ii ii  ! ii ii i11 I 0.9 .1 III IJ'I Ii 0.8

. ii.ii i i ! !i ! 1 1 1 .. ý . ýý i ii iii i i i ii i! ! i i I II I i i i i i i I ! i i i J i i i i ! i i i ! i i ! i I I I I I I I I I I i i i i ! i i ! i i i ! i i ! i i i i i i i i i i i ; ! i i ! i i i I I s0.7 U

I .

0.6 Core Elev. (fi) FQ K(Z) 0.0 2.55 1.0 7.0 2.55 1.0I 0.5

+ 14.0 2.359 0.925 0A 0.4 0- 1 1 1 1 1 1 1 1 1 1 1 1 1i 1 1 11 1 1 1 1 1 1 1 1 I 1 1 1 1 1 1 1 1 [ 1 1 i l[I Ii 1111 I1 M 1 1 1 1 1 1 1 1 1 M 1 1 1 1 I l1 1 II2 ------

I..' II I I I.1') 1 1 1 1 1 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

APATFAM Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 15 of 16 Table 1 (Part 1 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Less Than 10000 MWD/MTTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy I(Ft.) Point Fxy 14.00 1 5.702 6.80 37 1.969 13.80 2 4.737 6.60 38 1.965 13.60 3 3.771 6.40 39 1.960 13.40 4 2.806 6.20 40 1.958 13.20 5 2.360 6.00 41 1.957 13.00 6 2.146 5.80 42 1.959 12.80 7 2.098 5.60 43 1.962 12.60 8 2.079 5.40 44 1.966 12.40 9 2.061 5.20 45 1.971 12.20 10 2.036 5.00 46 1.974 12.00 11 2.011 4.80 47 1.977 11.80 12 1.995 4.60 48 1.977 11.60 13 1.993 4.40 49 1.979 11.40 14 1.987 4.20 50 1.981 11.20 15 1.975 4.00 51 1.982 11.00 16 1.957 3.80 52 1.976 10.80 17 1.947 3.60 53 1.968 10.60 18 1.941 3.40 54 1.961 10.40 19 1.939 3.20 55 1.957 10.20 20 1.944 3.00 56 1.954 10.00 21 1.958 2.80 57 1L951 9.80 22 1.978 2.60 58 1.951 9.60 23 1.995 2.40 59 1.946 9.40 24 2.009 2.20 60 1.934 9.20 25 2.021 2.00 61 1.882 9.00 26 2.033 1.80 62 1.818 8.80 27 2.046 1.60 63 1.749 8.60 28 2.060 1.40 64 1.749 8.40 29 2.075 1.20 65 1.760 8.20 30 2.086 1.00 66 1.794 8.00 31 2.079 0.80 67 1.928 7.80 32 2.061 0.60 68 2.136 7.60 33, 2.029 0.40 69 2.379 7.40 34 2.010 0.20 70 2.623 7.20 35 1.990 0.00 71 2.866 7.00 36 1.978

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 16 of.16 Table 1 (Part 2 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 10000 MW.D/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) , Point Fxy 14.00 1 4.313 6.80 37 2.144 13.80 2 3.817 6.60 38 2.149 13.60 3 3.321 6.40 .39 2.145 13.40 4 2.826 6.20 40 2.138 13.20 5 2.507 6.00 41 2.128 13.00 6 2.292 5.80 42 2.115 12.80 7 2.186 5.60 43 2.102 12.60 8 2.124 5.40 44 2.088 12.40 9 2.078 5.20 45 2.076 12.20 10 2.032 5.00 46 2.065 12.00 11 2.011 4.80 47 2.055 11.80 12 2.009 4.60 48 2.045 11.60 13 2.006 4.40 49 2.034 11.40 14 2.008 4.20 50 2.023 11.20 15 2.009 4.00 51 2.010 11.00 16 2.012 3.80 52 1.998 10.80 17 2.015 3.60 53 1.987 10.60 18 2.018 3.40 54 1.976 10.40 19 2.023 3.20 55 1.961 10.20 20 2.040 3.00 56 1.942 10.00 21 2.063 2.80 57 1.919 9.80 22 2.087 2.60 58 1.892 9.60 23 2.104 2.40 '59 1.863 9.40 24 2.112 2.20 60 1.835 9.20 25 2.114 2.00 61 1.824 9.00 26 2.111 1.80 62 1.817 8.80 27 2.107 1.60 63 1.815 8.60 28 2.103 1.40 64 1.814 8.40 29 2.101 1.20 65 1.848 8.20 30 2.100 1.00 66 1.928 8.00 31 2.100 0.80 67 2.167 7.80 32 2.101 0.60 68 2.588 7.60 33 2.107 0.40 69 3.099 7.40 34 2.115 0.20 70 3.610 7.20 35 2.125 0.00 71 4.122 7.00 36 2.136