ML15314A658

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Cycle 21 Core Operating Limits Report, Revision 1
ML15314A658
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 11/04/2015
From: Hafenstine C
Wolf Creek
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 15-0081
Download: ML15314A658 (17)
v • d • e


Text

LFCREEK NIUCLEAR OPERATING CORPORATION November 4, 2015 Cynthia R. Hafenstine Manager Regulatory Affairs RA 15-0081 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 21 Core Operating Limits Report, Revision 1 Gentlemen:

Enclosed is Revision 1 of the Wolf Creek Generating Station Cycle 21 Core Operating Limits Report (COLR). Revision 1 incorporates changes associated with the implementation of Amendment No. 213. This document is being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station Technical Specifications.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204.

Sincerely, Cynthia R. Hafenstine CRH/rlt Enclosure cc: M. L. Dapas (NRC), wle C. F. Lyon (NRC), w/e N. H. Taylor (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCINET

W@LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 WOLF CREEK GENERATING STATION CYCLE 21 CORE OPERATING LIMITS REPORT Revision 1 September 2015

  • /'* *"*9/8/15 Prepared by:

Jeff Blair Date Reviewed by:

Keith Colussy Date Approved by:

Gregory S. Kinn Date Page 1 of 16

W NULFA CCPRR CRORTO ICI¢K EEK ~Wolf Creek Generating StatiOncce2 Core Operating Limits Report CORORAIONRevision

'NUCEAROPERTIN I 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 21 has been prepared in accordance with the requirements of Technical Specification 5.6.5.

The core operating limits that are included in the COLR affect the following Technical Specifications:

2.1.1 Reactor Core Safety Limits 3.1 .1 Shutdown Margin (SDM) 3.1.3 Moderator Temperature Coefficient (MTC) 3.1 .4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 PHYSICS TESTS Exceptions - MODE 2 3.2.1 Heat Flux Hot Channel Factor (FQ(z)) (Fo Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (Fr) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) 3.3.1 Reactor Trip System (RTS) Instrumentation 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.9.1 Boron Concentration The portions of the Technical Specification Bases affected by the report are listed below:

ASA B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 16

W0LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the subsections below:

2.1 Reactor Core Safety Limits (SL 2.1.1)

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits in Figure 2.1.

680 660 640 LI-620 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16

W'*,FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.2 Moderator Temperature Coefficient (MTC') (LCO 3.1.3, SR 3.1.3.2)

The MTC shall be less positive than the limit provided in Figure 2.2.

The MTC shall be less negative than -50 pcm/°F.

The 300 PPM MTC Surveillance limit is -41 pcm/°F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).

The 60 PPM MTC Surveillance limit is -46 pcm/°F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).

UNACCEPTABLE OPERATION 6.0, 70%

U

  • .6 I-z w

0_

ACCEPTABLE OPERATION a.

0 0 10 20 30 40 50 60 70 80 90 100

% of RATED THERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs.

THERMAL POWER (%)

Page 4 of 16

Wolf Creek Generating Station W@LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 2.3 Shutdown Bank Insertion Limits (LCO 3.1.5)

The shutdown banks shall be fully withdrawn (i.e., positioned within the interval of > 222 and < 231 steps withdrawn).

2.4 Control Bank Insertion Limits (LCO 3.1 .6)

The Control Bank insertion, sequence, and overlap limits are specified in Figure 2.4.

(FULLY WITHDRAWN) 220 r---- 1- I-- *-o - - F1-* I-*- __ -I* Z--

200 180 160 S

T E 140 P

S W 120 T

H 100 D

R A 80 W

N 60 40 20 0

0 20 40 60 80 100 (FULLY INSERTED) THERMAL POWER (Percent)

Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs.

THERMAL POWER (%) - Four Loop Operation Fully withdrawn shall be the condition where control banks are at a position within the interval of>Ž 222 and < 231 steps withdrawn.

Page 5 of 16

Wolf Creek Generating Station

'WSLF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 2.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)

Methodoloaqy) (LCO 3.2.3)

The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space is defined by Figure 2.5.

110 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION F

90 R

A T

E D 80 T

H ACCEPTABLE OPERATION E

R70 M

A L

P6 0 E

R 50

( -29 ,50 ) ( 24,50 )

40  ! I I I I I I I  ! I I I I I

-40 -30 -20 -10 0 10 20 30 40 AXIAL FLUX DIFFERENCE (% AI)

Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER (%)

Page 6 of 16

WNUCLFA WLF C~lli=KWolf CPR CRORTO EEK Core Operating Limits Report Creek Generating StatiOnyce2

'NUCEAROPERTIN CORORAIONRevision 1 2.6 Heat Flux Hot Channel Factor (FoIZY)(F 0 Methodologqy) (LCO 3.2.1, SR 3.2.1 .2)

Fo(Z)*_FQ-**K(Z), forP>O0.5 FQ(Z)<_C.Q5*K(Z), for P *_ 0.5 where

~ - THERMAL POWER RATED THERMAL POWER T

CFQ = Flr P F~P= Fo(Z) limit at RATED THERMAL POWER (RTP)

= 2.50, and K(Z) = as defined in Figure 2.6.

Fo'(Z) is the measured value of F0 (Z), inferred from a power distribution measurement obtained with the Movable Incore Detector System (MIDS) or the Power Distribution Monitoring System.

Measurement uncertainty is applied as follows.

FffZ)= FQA(Z)(l.O3)(l.O5)=F*l(Z)(1.0815) when FoM (Z) is obtained from MIDS.

Fg(Z) = F*(Z)(I.O3)(UQU) when FoM(Z) is obtained from PDMS.

Manufacturing tolerances are accounted for in the 1.03 Engineering uncertainty factor. Measurement uncertainty for MIDS is accounted for in the 1.05 factor.

PDMS measurement uncertainty is accounted for in the UQU factor, and it is determined by PDMS.

where, W(Z) = a cycle dependent function that accounts for power distribution transients encountered during normal operation (see Appendix A).

When using the PDMS, FoY(Z) uses F*(Z) that is determined from an F*(~Z) that reflects full-power steady-state conditions rather than current conditions.

See Appendix A for: FQ Penalty Factor.

Page 7 of 16

Wolf Creek Generating Station WCLF CREEK T NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 1.2 ir1.0 0.

C. 0.8 S0.

z1 06

+ +

z 0.2 FQRW= = 2.50 Bevation (ft) K(Z)-

0.0 1.000 6.0 1.000 12.0 0.925 0.0 l I 0 2 4 6 8 10 12 CORE HEIGHT (FT)

Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Height Page 8 of 16

WNULEAR WLF CRI¢ICKWolf CPR CRORTOEEK Creek Generating StatiOncce2 Core Operating Limits Report OPERTIN CORORAIONRevision I 2.7 Nuclear Enthalpy Rise Hot Channel Factor (Ft) (LCO 3.2.2)

F* shall be limited by the following relationship:

FNJ _<FLfft[1.O + PF*(1.O -P]

Where, F7p = F* limit at RATED THERMAL POWER (RTP)

= 1.650 PF* = power factor multiplier for F*

= 0.3 P = THERM/AL POWER RATED THERMAL POWER F[H = F* is the measured value of F*, inferred from a power distribution measurement obtained with the Movable Incore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS). Measurement uncertainty is applied as follows.

When F*j is obtained from MIDS, the measured value is multiplied by 1.04.

When F* is obtained from PDMS, the measured value is increased by an uncertainty factor (UAH), and the factor is determined by PDMS, with a lower limit of 4%.

Page 9 of 16

W6FCREEK PRTNG CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision I 2.8 Reactor Trip~ System Overtemperature AT Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 1)

Parameter Value K1 = 1.10 Overtemperature AT reactor trip setpoint Overtemperature AT reactor trip setpoint Tavg K2 = 0.01 37/°F coefficient Overtemperature AT reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal Tavg at RTP T' < 586.5°F Nominal RCS operating pressure P' >Ž2235 psig Measured RCS AT lead/lag constant 1= 6 sec 1t2 = 3 sec Measured RCS AT lag constant 13= 2 sec Measured RCS average temperature lead/lag "1 4 =16 sec MeasurdRtaergaeprauela/lg=te 15= 4 sec constant f1(AI) = -0.0227 {23% + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP < (qt-qb,) < 5% RTP 0.0184 {(qt-qb) - 5%} when (qt-qb,) > 5% RTP Where, qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

Page 10 of 16

W0LF CREEK rNUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 2)

Parameter Value Overpower AT reactor trip setpoint K4= 1.10 Overpower AT reactor trip setpoint Tavg I 5 =O.02/0 F for increasing Tavg rate/lag coefficient = 0/°F for decreasing Tavg Overpower AT reactor trip setpoint Tavg heatup I*= 0.001 28/°F for T > T" coefficient = 0/OF forT*<T,,

Indicated mavg at RTP (calibration temperature T" _<586.5°F for AT instrumentation)

Measured RCS AT lead/lag constant 11= 6 sec 1;2 = 3 sec Measured RCS AT lag constant 13= 2 sec Measured RCS average temperature lead/lag 16= 0 sec constant Measured RCS average temperature rate/lag  ; 7 = 10 sec constant f2(AI) = 0% RTP for all Al Page 11 of 16

'W 7

LF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling ('DNB)

Limits (LCO 3.4.1)

Parameter Indicated Value Pressurizer pressure Pressure Ž_2220 psig RCS average temperature Tavg*<590.5 °F RCS total flow rate Flow > 371,000 gpm 2.11 Boron Concentration (LCO 3.9.1)

The refueling boron concentration shall be greater than or equal to 2300 PPM.

2.12 SHUTDOWN MARGIN (LCO 3.1.1, 3.1.4, 3.1.5, 3.1.6, & 3.1.8)

The SHUTDOWN MARGIN shall be greater than or equal to 1300 pcm (1.3% Ak/k).

2.13 Departure from Nucleate Boilinq Ratio (DNBR) Limits (B 3.4.1, ASA)

Safety Analysis DNBR Limit 1.76 WRB-2 Design Limit DNBR 1.23 Page 12 of 16

W@LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 APPENDIX A A. Input relating to LCO 3.2.1:

W() Q( Z )steadystate for P > 0.5 W(z) 1 for P <_ 0.5 EQ (Z)stcadystate 0.5' wher~ = THERMAL POWER RATED THERMAL POWER FQ(Z)maxtransient = Maximum (FQ('Z) xp) calculated over the entire range of power shapes analyzed for Condition I operations (p = power at which maximum occurs).

FQ(Z)stea'stae = (FQ(Z) xp) calculated at full power (p = 1.0) equilibrium conditions.

The W(z) values are generated at full power equilibrium conditions (P = 1.0). W(z) values specific to part-power conditions may also be generated; these can be used for part-power surveillance measurements, rather than the full-power W(z) values. For these part-power W(z) values, the FQ(Z)steady state (denominator in above equations) is generated at the specific anticipated surveillance conditions.

W(Z) values are issued in controlled reports which will be provided on request.

Page 13 of 16

Wolf Creek Generating Station W#LF

  • CREEK

'NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 Input relating to SR 3.2.1.2 Cycle Burnup F0o(Z) Penalty Factor (MWD/MTU) (%)

> 0 to <*150 3.08 348 2.76 546 2.38 743 to 8456 2.00 8654 2.06 8852 2.26 9049 2.44 9247 2.40 9445 2.38 9643 2.36 9840 2.31 10038 2.23 10236 2.15 10434 2.06

> 10631 2.00 Fo(7) Exclusion Zone

(% [INCORE mesh Cycle Burnup points])

(MWD/MTU) Top I:Bottom

<88000

> 8,000 10 [7] I: 10 [7]

Page 14 of 16

WUCLFA CREEK CPR EEK CRORTO

'NUCEAROPERTIN CORORAIONRevision

~ ~Wolf Creek yleGenerating 21Station Core Operating Limits Report 1

B. App~roved Analytical Methods for Determining Core Op~eratinaq Limits The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

1. WCNOC Topical Report TR 90-0025 WO1, "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station." (ET 90-0140, ET 92-0103)

NRC Safety Evaluation Report dated October 29, 1992, for the "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station."

2. WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.

NRC Safety Evaluation Report dated January 17, 1989, for the "Acceptance for Referencing of Licensing Topical Report WCAP-1 1397, Revised Thermal Design Procedure."

3. WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the Wolf Creek Generating Station" (ET-91-0026, ET 92-0142, WM 93-001 0, WM~ 93-0028).

NRC Safety Evaluation Report dated September 30, 1993, for the "Transient Analysis Methodology for the Wolf Creek Generating ,Station."

EPRI Topical Report NP-7450(A), "RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," including NRC Safety Evaluation Report dated January 25, 2001, "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D -A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,"

(TAC No. MA431 1)." RETRAN-3D code is only utilized in the RETRAN-02 mode.

4. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.

NRC Safety Evaluation Report dated November 26, 1993, "Acceptance for Referencing of Revised Version of Licensing Topical Report WCAP-1 021 6-P, Rev. 1, Relaxation of Constant Axial Offset Control - F0 Surveillance Technical ,Specification" (TAC No. M88206).

5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017).

NRC ,Safety Evaluation Report dated March 10, 1993, for the "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."

6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054).

Page 15 of 16

.....C R I KWolf Creek Generating Station

'NUCLEAR OPERATING CORPORATION RoeO evisiong LimisRpr

7. WCAP-1 6009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

Revision 0, January 2005.

NRC letter dated November 5, 2004,"Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)"

(TAC NO. MB9483)."

8. WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004.

NRC Safety Evaluation dated March 18, 2004, "Final Safety Evaluation for Westinghouse Topical Report WCAP-1 6045-P, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON."

9. WCAP-1 6045-P-A, Addendum I-A, "Qualification of the NEXUS Nuclear Data Methodology," August 2007.

NRC Safety Evaluation dated February 23, 2007, "Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16045-P-A, Addendum 1, "Qualification of the NEXUS Nuclear Data Methodology" (TAC NO. MC9606)."

10. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"

September 1986.

NRC letter dated June 23, 1986, "Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP."

11. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.

NRC Safety Evaluation Reports dated July 1, 1991, "Acceptance for Referencing of Topical Report WCAP-1 2610, 'VANTAGE+ Fuel Assembly Reference Core Report' (TAC NO. 77258)."

NRC Safety Evaluation Report dated September 15, 1994, "Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1,

'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M8641 6)."

12. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Function." September 1986.

NRC Safety Evaluation Report dated April 17, 1986, "Acceptance for Referencing of Licensing Topical Report WCAP-8745(P)/8746(NP), 'Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions."'"

Page 16 of 16

LFCREEK NIUCLEAR OPERATING CORPORATION November 4, 2015 Cynthia R. Hafenstine Manager Regulatory Affairs RA 15-0081 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 21 Core Operating Limits Report, Revision 1 Gentlemen:

Enclosed is Revision 1 of the Wolf Creek Generating Station Cycle 21 Core Operating Limits Report (COLR). Revision 1 incorporates changes associated with the implementation of Amendment No. 213. This document is being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station Technical Specifications.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204.

Sincerely, Cynthia R. Hafenstine CRH/rlt Enclosure cc: M. L. Dapas (NRC), wle C. F. Lyon (NRC), w/e N. H. Taylor (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCINET

W@LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 WOLF CREEK GENERATING STATION CYCLE 21 CORE OPERATING LIMITS REPORT Revision 1 September 2015

  • /'* *"*9/8/15 Prepared by:

Jeff Blair Date Reviewed by:

Keith Colussy Date Approved by:

Gregory S. Kinn Date Page 1 of 16

W NULFA CCPRR CRORTO ICI¢K EEK ~Wolf Creek Generating StatiOncce2 Core Operating Limits Report CORORAIONRevision

'NUCEAROPERTIN I 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 21 has been prepared in accordance with the requirements of Technical Specification 5.6.5.

The core operating limits that are included in the COLR affect the following Technical Specifications:

2.1.1 Reactor Core Safety Limits 3.1 .1 Shutdown Margin (SDM) 3.1.3 Moderator Temperature Coefficient (MTC) 3.1 .4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 PHYSICS TESTS Exceptions - MODE 2 3.2.1 Heat Flux Hot Channel Factor (FQ(z)) (Fo Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (Fr) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) 3.3.1 Reactor Trip System (RTS) Instrumentation 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.9.1 Boron Concentration The portions of the Technical Specification Bases affected by the report are listed below:

ASA B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 16

W0LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the subsections below:

2.1 Reactor Core Safety Limits (SL 2.1.1)

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits in Figure 2.1.

680 660 640 LI-620 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16

W'*,FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.2 Moderator Temperature Coefficient (MTC') (LCO 3.1.3, SR 3.1.3.2)

The MTC shall be less positive than the limit provided in Figure 2.2.

The MTC shall be less negative than -50 pcm/°F.

The 300 PPM MTC Surveillance limit is -41 pcm/°F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).

The 60 PPM MTC Surveillance limit is -46 pcm/°F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).

UNACCEPTABLE OPERATION 6.0, 70%

U

  • .6 I-z w

0_

ACCEPTABLE OPERATION a.

0 0 10 20 30 40 50 60 70 80 90 100

% of RATED THERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs.

THERMAL POWER (%)

Page 4 of 16

Wolf Creek Generating Station W@LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 2.3 Shutdown Bank Insertion Limits (LCO 3.1.5)

The shutdown banks shall be fully withdrawn (i.e., positioned within the interval of > 222 and < 231 steps withdrawn).

2.4 Control Bank Insertion Limits (LCO 3.1 .6)

The Control Bank insertion, sequence, and overlap limits are specified in Figure 2.4.

(FULLY WITHDRAWN) 220 r---- 1- I-- *-o - - F1-* I-*- __ -I* Z--

200 180 160 S

T E 140 P

S W 120 T

H 100 D

R A 80 W

N 60 40 20 0

0 20 40 60 80 100 (FULLY INSERTED) THERMAL POWER (Percent)

Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs.

THERMAL POWER (%) - Four Loop Operation Fully withdrawn shall be the condition where control banks are at a position within the interval of>Ž 222 and < 231 steps withdrawn.

Page 5 of 16

Wolf Creek Generating Station

'WSLF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 2.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)

Methodoloaqy) (LCO 3.2.3)

The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space is defined by Figure 2.5.

110 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION F

90 R

A T

E D 80 T

H ACCEPTABLE OPERATION E

R70 M

A L

P6 0 E

R 50

( -29 ,50 ) ( 24,50 )

40  ! I I I I I I I  ! I I I I I

-40 -30 -20 -10 0 10 20 30 40 AXIAL FLUX DIFFERENCE (% AI)

Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER (%)

Page 6 of 16

WNUCLFA WLF C~lli=KWolf CPR CRORTO EEK Core Operating Limits Report Creek Generating StatiOnyce2

'NUCEAROPERTIN CORORAIONRevision 1 2.6 Heat Flux Hot Channel Factor (FoIZY)(F 0 Methodologqy) (LCO 3.2.1, SR 3.2.1 .2)

Fo(Z)*_FQ-**K(Z), forP>O0.5 FQ(Z)<_C.Q5*K(Z), for P *_ 0.5 where

~ - THERMAL POWER RATED THERMAL POWER T

CFQ = Flr P F~P= Fo(Z) limit at RATED THERMAL POWER (RTP)

= 2.50, and K(Z) = as defined in Figure 2.6.

Fo'(Z) is the measured value of F0 (Z), inferred from a power distribution measurement obtained with the Movable Incore Detector System (MIDS) or the Power Distribution Monitoring System.

Measurement uncertainty is applied as follows.

FffZ)= FQA(Z)(l.O3)(l.O5)=F*l(Z)(1.0815) when FoM (Z) is obtained from MIDS.

Fg(Z) = F*(Z)(I.O3)(UQU) when FoM(Z) is obtained from PDMS.

Manufacturing tolerances are accounted for in the 1.03 Engineering uncertainty factor. Measurement uncertainty for MIDS is accounted for in the 1.05 factor.

PDMS measurement uncertainty is accounted for in the UQU factor, and it is determined by PDMS.

where, W(Z) = a cycle dependent function that accounts for power distribution transients encountered during normal operation (see Appendix A).

When using the PDMS, FoY(Z) uses F*(Z) that is determined from an F*(~Z) that reflects full-power steady-state conditions rather than current conditions.

See Appendix A for: FQ Penalty Factor.

Page 7 of 16

Wolf Creek Generating Station WCLF CREEK T NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 1.2 ir1.0 0.

C. 0.8 S0.

z1 06

+ +

z 0.2 FQRW= = 2.50 Bevation (ft) K(Z)-

0.0 1.000 6.0 1.000 12.0 0.925 0.0 l I 0 2 4 6 8 10 12 CORE HEIGHT (FT)

Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Height Page 8 of 16

WNULEAR WLF CRI¢ICKWolf CPR CRORTOEEK Creek Generating StatiOncce2 Core Operating Limits Report OPERTIN CORORAIONRevision I 2.7 Nuclear Enthalpy Rise Hot Channel Factor (Ft) (LCO 3.2.2)

F* shall be limited by the following relationship:

FNJ _<FLfft[1.O + PF*(1.O -P]

Where, F7p = F* limit at RATED THERMAL POWER (RTP)

= 1.650 PF* = power factor multiplier for F*

= 0.3 P = THERM/AL POWER RATED THERMAL POWER F[H = F* is the measured value of F*, inferred from a power distribution measurement obtained with the Movable Incore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS). Measurement uncertainty is applied as follows.

When F*j is obtained from MIDS, the measured value is multiplied by 1.04.

When F* is obtained from PDMS, the measured value is increased by an uncertainty factor (UAH), and the factor is determined by PDMS, with a lower limit of 4%.

Page 9 of 16

W6FCREEK PRTNG CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision I 2.8 Reactor Trip~ System Overtemperature AT Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 1)

Parameter Value K1 = 1.10 Overtemperature AT reactor trip setpoint Overtemperature AT reactor trip setpoint Tavg K2 = 0.01 37/°F coefficient Overtemperature AT reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal Tavg at RTP T' < 586.5°F Nominal RCS operating pressure P' >Ž2235 psig Measured RCS AT lead/lag constant 1= 6 sec 1t2 = 3 sec Measured RCS AT lag constant 13= 2 sec Measured RCS average temperature lead/lag "1 4 =16 sec MeasurdRtaergaeprauela/lg=te 15= 4 sec constant f1(AI) = -0.0227 {23% + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP < (qt-qb,) < 5% RTP 0.0184 {(qt-qb) - 5%} when (qt-qb,) > 5% RTP Where, qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

Page 10 of 16

W0LF CREEK rNUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 2)

Parameter Value Overpower AT reactor trip setpoint K4= 1.10 Overpower AT reactor trip setpoint Tavg I 5 =O.02/0 F for increasing Tavg rate/lag coefficient = 0/°F for decreasing Tavg Overpower AT reactor trip setpoint Tavg heatup I*= 0.001 28/°F for T > T" coefficient = 0/OF forT*<T,,

Indicated mavg at RTP (calibration temperature T" _<586.5°F for AT instrumentation)

Measured RCS AT lead/lag constant 11= 6 sec 1;2 = 3 sec Measured RCS AT lag constant 13= 2 sec Measured RCS average temperature lead/lag 16= 0 sec constant Measured RCS average temperature rate/lag  ; 7 = 10 sec constant f2(AI) = 0% RTP for all Al Page 11 of 16

'W 7

LF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 2.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling ('DNB)

Limits (LCO 3.4.1)

Parameter Indicated Value Pressurizer pressure Pressure Ž_2220 psig RCS average temperature Tavg*<590.5 °F RCS total flow rate Flow > 371,000 gpm 2.11 Boron Concentration (LCO 3.9.1)

The refueling boron concentration shall be greater than or equal to 2300 PPM.

2.12 SHUTDOWN MARGIN (LCO 3.1.1, 3.1.4, 3.1.5, 3.1.6, & 3.1.8)

The SHUTDOWN MARGIN shall be greater than or equal to 1300 pcm (1.3% Ak/k).

2.13 Departure from Nucleate Boilinq Ratio (DNBR) Limits (B 3.4.1, ASA)

Safety Analysis DNBR Limit 1.76 WRB-2 Design Limit DNBR 1.23 Page 12 of 16

W@LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 21 Core Operating Limits Report Revision 1 APPENDIX A A. Input relating to LCO 3.2.1:

W() Q( Z )steadystate for P > 0.5 W(z) 1 for P <_ 0.5 EQ (Z)stcadystate 0.5' wher~ = THERMAL POWER RATED THERMAL POWER FQ(Z)maxtransient = Maximum (FQ('Z) xp) calculated over the entire range of power shapes analyzed for Condition I operations (p = power at which maximum occurs).

FQ(Z)stea'stae = (FQ(Z) xp) calculated at full power (p = 1.0) equilibrium conditions.

The W(z) values are generated at full power equilibrium conditions (P = 1.0). W(z) values specific to part-power conditions may also be generated; these can be used for part-power surveillance measurements, rather than the full-power W(z) values. For these part-power W(z) values, the FQ(Z)steady state (denominator in above equations) is generated at the specific anticipated surveillance conditions.

W(Z) values are issued in controlled reports which will be provided on request.

Page 13 of 16

Wolf Creek Generating Station W#LF

  • CREEK

'NUCLEAR OPERATING CORPORATION Cycle 21 Core Operating Limits Report Revision 1 Input relating to SR 3.2.1.2 Cycle Burnup F0o(Z) Penalty Factor (MWD/MTU) (%)

> 0 to <*150 3.08 348 2.76 546 2.38 743 to 8456 2.00 8654 2.06 8852 2.26 9049 2.44 9247 2.40 9445 2.38 9643 2.36 9840 2.31 10038 2.23 10236 2.15 10434 2.06

> 10631 2.00 Fo(7) Exclusion Zone

(% [INCORE mesh Cycle Burnup points])

(MWD/MTU) Top I:Bottom

<88000

> 8,000 10 [7] I: 10 [7]

Page 14 of 16

WUCLFA CREEK CPR EEK CRORTO

'NUCEAROPERTIN CORORAIONRevision

~ ~Wolf Creek yleGenerating 21Station Core Operating Limits Report 1

B. App~roved Analytical Methods for Determining Core Op~eratinaq Limits The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

1. WCNOC Topical Report TR 90-0025 WO1, "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station." (ET 90-0140, ET 92-0103)

NRC Safety Evaluation Report dated October 29, 1992, for the "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station."

2. WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.

NRC Safety Evaluation Report dated January 17, 1989, for the "Acceptance for Referencing of Licensing Topical Report WCAP-1 1397, Revised Thermal Design Procedure."

3. WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the Wolf Creek Generating Station" (ET-91-0026, ET 92-0142, WM 93-001 0, WM~ 93-0028).

NRC Safety Evaluation Report dated September 30, 1993, for the "Transient Analysis Methodology for the Wolf Creek Generating ,Station."

EPRI Topical Report NP-7450(A), "RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," including NRC Safety Evaluation Report dated January 25, 2001, "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D -A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,"

(TAC No. MA431 1)." RETRAN-3D code is only utilized in the RETRAN-02 mode.

4. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.

NRC Safety Evaluation Report dated November 26, 1993, "Acceptance for Referencing of Revised Version of Licensing Topical Report WCAP-1 021 6-P, Rev. 1, Relaxation of Constant Axial Offset Control - F0 Surveillance Technical ,Specification" (TAC No. M88206).

5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017).

NRC ,Safety Evaluation Report dated March 10, 1993, for the "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."

6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054).

Page 15 of 16

.....C R I KWolf Creek Generating Station

'NUCLEAR OPERATING CORPORATION RoeO evisiong LimisRpr

7. WCAP-1 6009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

Revision 0, January 2005.

NRC letter dated November 5, 2004,"Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)"

(TAC NO. MB9483)."

8. WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004.

NRC Safety Evaluation dated March 18, 2004, "Final Safety Evaluation for Westinghouse Topical Report WCAP-1 6045-P, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON."

9. WCAP-1 6045-P-A, Addendum I-A, "Qualification of the NEXUS Nuclear Data Methodology," August 2007.

NRC Safety Evaluation dated February 23, 2007, "Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16045-P-A, Addendum 1, "Qualification of the NEXUS Nuclear Data Methodology" (TAC NO. MC9606)."

10. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"

September 1986.

NRC letter dated June 23, 1986, "Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP."

11. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.

NRC Safety Evaluation Reports dated July 1, 1991, "Acceptance for Referencing of Topical Report WCAP-1 2610, 'VANTAGE+ Fuel Assembly Reference Core Report' (TAC NO. 77258)."

NRC Safety Evaluation Report dated September 15, 1994, "Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1,

'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M8641 6)."

12. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Function." September 1986.

NRC Safety Evaluation Report dated April 17, 1986, "Acceptance for Referencing of Licensing Topical Report WCAP-8745(P)/8746(NP), 'Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions."'"

Page 16 of 16

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