ML130720680

Cycle 20 Core Operating Limits Report

ML130720680
Person / Time
Site:	Wolf Creek
Issue date:	03/04/2013
From:	Westman M Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 13-0027
Download: ML130720680 (17)

Text

WOLF CREEK NUCLEAR OPERATING CORPORATION Michael J. Westman Manager Regulatory Affairs March 4, 2013 RA 13-0027 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 20 Core Operating Limits Report Gentlemen:

Enclosed is Revision 0 of the Wolf Creek Generating Station Cycle 20 Core Operating Limits Report (COLR). 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) 364-8831 ext. 4009.

Sincerely, Michael J. Westman MJW/rlt Enclosure cc: E. E. Collins (NRC), w/e C. F. Lyon (NRC), wle N. F. O'Keefe (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 Aco I An Equal Opportunity Employer M/F/HCNET

Enclosure to RA 13-0027 WOLF CREEK GENERATING STATION CYCLE 20 CORE OPERATING LIMITS REPORT, Revision 0 (15 pages)

Wolf Creek Generating Station W#4LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 20 CORE OPERATING LIMITS REPORT Revision 0 January 2013 01/23/2013 Prepared by: ,ro-ý 74. (ý-

James H. Eaves Date

• " * *1/28/13 Reviewed by:

Jeff Blair Date Approved by: JWY51.,i2/24/13 Gregory S. Kinn Date Page 1 of 15

j CWolf Creek Generating Station We CREEK OPLF

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 20 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)) (FQ Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (k".)

3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) 3.3.1 Reactor Trip System (RTS) Instrumentation 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

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 15

Wolf Creek Generating Station W*4LF CREEK

'NUCLEAR OPERATING CORPORATION Core Cycle 20 Operating Limits Report Revision 0 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.

680 660 640 620

.2 0

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 15

Wolf Creek Generating Station W4LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 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).

8 z

L6 0

I-1.

12 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 15

Wolf Creek Generating Station W*ALF CREEK

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Cycle 20 Revision 0 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 I

,,- , -. . . .- ,2 2 2 )

F Fi F F F i F F F 200 i i F F I F F i 180 -- ------ ..-------- H- ---.------

- - -AN - -- - ---------

-- --H

, ( 1000/0161 )

160 S

T - -- F- - -- - - K E 140 30.12%

F FFF F F F F 0

• ~ F F P r F FA'I F F F F F F F F F F S F F F F F FF F Fi

- F F FF F F F F Fi F F F F F F F F Ii W 120 F F I F F FF F Ii 1 --- -i -t---+

T H 100 D -- Ll - ----- -- - - - -*-

R F F F F F F FF F , BAN ,K F A 80 W

N 60 F F I Fi F F F F i F 40 F FiFi F F F F F F F

.. b-A- - - -- --- - - - - -b Fi i F F F F F F F F F]/ 46 02 0/!, F, .F F. F F F 20 0

0 20 40 60 80 100 (FULLY INSERTED) THERMAL POW ER (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 > 222 and < 231 steps withdrawn.

Page 5 of 15

Wolf Creek Generating Station WejLF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 2.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)

Methodoloqy) (LCO 3.2.3)

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

110

(-15 ,100 ) 5 ,100) 100 - UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 0

F 90 R

A T

E D 80 T

ACCEPTABLE H OPERATION E

R 70 M

A L

P 60 0

W E

R 50

(-29 ,50) (24 ,50) 40

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

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

Page 6 of 15

Wolf Creek Generating Station W*4LF CREEK 1 111111111K Cycle 20

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (FQ(Z))(FO Methodology) (LCO 3.2.1, SR 3.2.1.2)

FL(Z) CFQ *K(Z), for P > 0.5 FQ(Z)<EF-Q *K(Z), forP

• 0.5 0.5 where, P - THERMAL POWER RATED THERMAL POWER CFQ = F F RI = FQ(Z) limit at RATED THERMAL POWER (RTP)

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

FQ'(Z) is the measured value of FQ(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.

F*(Z) = F*(Z)(1.03)(1.05)=FQ m (Z)(1.0815) when FQm (Z) is obtained from MIDS.

Fo(Z) = F*m(Z)(1.03)(ULu) when FQ1 (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.

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

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

See Appendix A for: F. Penalty Factor.

Page 7 of 15

Wolf Creek Generating Station W4LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 1.2 NA 1.0 0

I-IZt., 0.8 O

Z w 0.6 a-0 Lu

_N

.-J 0.4 z 0.2 0.0 0 2 4 6 8 10 12 CORE HEIGHT (FT)

Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Height Page 8 of 15

Wolf Creek Generating Station W*4LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor (Fm,) (LCO 3.2.2)

FN, shall be limited by the following relationship:

F'_< t " PFAH (I"-P)

Where, FHRTP = FN limit at RATED THERMAL POWER (RTP)

= 1.650 PFAH= power factor multiplier for FN

= 0.3 P = THERMAL POWER RATED THERMAL POWER FN = FN is the measured value of FN, 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', is obtained from MIDS, the measured value is multiplied by 1.04.

When FA 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 15

Wolf Creek Generating Station W*JLF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 2.8 Reactor Trip System Overtemperature AT Setpoint Parameter Values (LCO 3.3.1)

Parameter Value Overtemperature AT reactor trip setpoint K1 = 1.10 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 0 F Nominal RCS operating pressure P' > 2235 psig Measured RCS AT lead/lag constant Tri = 6 sec T2 = 3 sec Measured RCS AT lag constant T3-= 2 sec Measured RCS average temperature lead/lag T4 =16 sec constant T5 =4 sec Measured RCS average temperature lead/lag T6 = 0 sec constant fd(Al) = -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 15

W LWolf Creek Generating Station W*4LF CREEK

'NUCLEAR OPERATING CORPORATION CoreOertn Liit0epr Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.2)

Parameter Value Overpower AT reactor trip setpoint K4 = 1.10 Overpower AT reactor trip setpoint Tavg K5 = 0.02/1F for increasing Tavg rate/lag coefficient = 0/°F for decreasing Tavg Overpower AT reactor trip setpoint Tavg heatup K6 = 0.00128/°F for T> T" coefficient = 0/°F for T:_T" Indicated Tavg at RTP (calibration temperature T" < 586.5 0 F for AT instrumentation)

Measured RCS AT lead/lag constant r, = 6 sec T2 = 3 sec Measured RCS AT lag constant T3 = 2 sec Measured RCS average temperature lead/lag T6 = 0 sec constant Measured RCS average temperature rate/lag -17 = 10 sec constant f2 (AI) = 0% RTP for all Al Page 11 of 15

Wolf Creek Generating Station W'ILF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 2.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits (LCO 3.4.1)

Parameter Indicated Value Pressurizer pressure Pressure >_2220 psig RCS average temperature Tavg < 590.5 OF 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 Boiling Ratio (DNBR) Limits (B 3.4.1, ASA)

Safety Analysis DNBR Limit 1.76 WRB-2 Design Limit DNBR 1.23 Page 12 of 15

Wolf Creek Generating Station W LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 20 Core Operating Limits Report Revision 0 APPENDIX A A. Input relating to LCO 3.2.1:

(Z)ax transient 1

W(z) - FQ FQ (Z)steadY stae x-,t p for P > 0.5 W(Z) = (Z)max transient for P _<0.5 FQ (Z)

St-mYS'1 0.5 where, P = THERMAL POWER RATED THERMAL POWER FQ(Z)max transient = Maximum (FQ(Z) xp) calculated over the entire range of power shapes analyzed for Condition I operations (p = power at which maximum occurs).

F,(Z)steastate = (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.

Input relating to SR 3.2.1.2 Cycle Burnup FeQ (Z) Penalty Factor (MWD/MTU) (%)

> 0 to < 621 2.10

> 621 2.00 FQ (Z) Exclusion Zone Cycle Burnup (%)

(MWD/MTU) Top Bottom

< 8,000 15 15

> 8,000 to --16,000 10 10

> 16,000 8 8 Page 13 of 15

WL- Wolf Creek Generating Station W 0LF CREEK Core Operating Limits Report ROPERATING CORPORATION Revision 0 B. Approved Analytical Methods for Determining Core Operatinq 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 W01, "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 93-0010, 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. MA4311)." 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-10216-P, Rev. 1, Relaxation of Constant Axial Offset Control - FQ 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).

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Wolf Creek Generating Station NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0

7. WCAP-10266-P-A, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.

NRC letter dated November 13, 1986, "Acceptance for Referencing of Licensing Topical Report WCAP-10266 "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code.""

WCAP-10266-P-A, Addendum 1, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code Addendum 1:

Power Shape Sensitivity Studies," December 1987.

NRC letter dated September 15, 1987, "Acceptance for Referencing of Addendum 1 to WCAP-1 0266, BASH Power Shape Sensitivity Studies."

WCAP-10266-P-A, Addendum 2, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code Addendum 2:

BASH Methodology Improvements and Reliability Enhancements," May 1988 NRC letter dated January 20, 1988, "Acceptance for Referencing Topical Report Addendum 2 to WCAP-1 0266, Revision 2, "BASH Methodology Improvements and Reliability Enhancements."

8. WCAP-1 1596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.

NRC Safety Evaluation Report dated May 17, 1988, "Acceptance for Referencing of Westinghouse Topical Report WCAP-1 1596 - Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."

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

September 1988.

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

10. 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-12610, '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. M86416)."

11. 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."'

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