ET 08-0031, Cycle 17 Core Operating Limits Report

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Cycle 17 Core Operating Limits Report
ML081440413
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 05/10/2008
From: Garrett T
Wolf Creek
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 08-0031
Download: ML081440413 (17)


Text

UCLEAR OPERATING CORPORATION Terry J. Garrett Vice President, Engineering May 10, 2008 ET 08-0031 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Gentlemen:

Enclosed is Revision 0 of the Wolf Creek Generating Station Cycle 17 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) 364-4084, or Mr. Richard D. Flannigan at (620) 364-4117.

Sincerely, Terry J. Garrett TJG/rlt Enclosure cc:

E. E. Collins (NRC), w/e V. G. Gaddy (NRC), w/e B. K. Singal (NRC), w/e Senior Resident Inspector (NRC), wle P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET 4OcD (

Enclosure to ET 08-0031 Wolf Creek Generating Station Cycle 17 - Core Operating Limits Report, Rev. 0

W*LF CREEK

'NUCLEARI OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 17 CORE OPERATING LIMITS REPORT Revision 0 March 2008 Prepared by:

Reviewed by:

Approved by:

02/14/2008 Matthew K. Morris Date o:;ý 3/3/2008 Jeff T. Blair Date 03/04/08 William S. Kennamore Date Page 1 of 15

Wolf Creek Generating Station Wý L F CREEKCycle 17 WCLFR CR EG CCore Operating Limits Report

'NUCLEAR OPERATING CORPORATION Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 17 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 (IL) 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

W;0LF CREEK F NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core 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

4) 620 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 Fraction of Rated Thermal Power 1.2 Figure 2.1 Reactor Core Safety Limits Page 3 of 15

W@LF CREEK INUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 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/OF (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).

8 E

z L6 U'

0 I.-

19 0

0 UNACCEPTABLE OPERATION 6.0, 70%

ACCEPTABLE OPERATION 0

10 20 30 40 50 60

% of RATED THERMAL POWER 70 80 90 100 Figure 2.2 Moderator Temperature Coefficient Vs.

RATED THERMAL POWER Page 4 of 15

W LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report 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 S

T E

P S

W I

T H

D R

A W

N 200 180 160 140 120 100 80 60

, 2

-ro -,222) i 76.7%

I I

I B

I I

II r

I I

I I

I I

I I

T - - - - - -.

I I

I I

I I

I I

I F

T I

I I

I I

I I

-R I

I I

I I

I I

I I

I I

I I

III I

I I

I F

I III II I

I I.

T FT------

F (0%,4'6 I

I

-I I

I I

I I

I I

. J.

. I.

F

.I L

.I.

J.

222 )

%! 161 )

I I

40 20 0

0 20 40 60 (FULLY INSERTED)

RATED THERMAL POWER 80 (P e rc e n t) 100 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

WCREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 2.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)

Methodology) (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

H E

R 70 M

A L

P 60 0

W E

R 50 40

-29 60 ACCEPTABLE OPERATION

(-29,50(24.50

-40

-30

-20

-10 0

10 20 AXIAL FLUX DIFFERENCE (%Al)

Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of RATED THERMAL POWER 30 40 Page 6 of 15

W6LF CREEK FNUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (Fo(Z))(FQ Methodology) (LCO 3.2.1, SR 3.2.1.2)

FQ(Z)< -FQ*K(Z), for P > 0.5 P

FQ(Z)< CFQ*K(Z), forP

  • 0.5 0.5 THERMAL POWER where, P RATED THERMAL POWER CFQ = FQrP FRTP = FQ(Z) limit at RATED THERMAL POWER (RTP)

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

F§ (Z) = FQj (Z)(1.03)(1.05) = FQ" (Z)(1.0815) where, FQj(Z) = Measured value of FQ(Z) from incore flux map Fw (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).

See Appendix A for:

Fo Penalty Factor Page 7 of 15

W*LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 1.2 N

0 L-C.)

Z U-0w 0

zU 0Z 1.0 0.8 0.6 0.4 0.2

-~----------I------

F RIP' 2.50 Elevation (ft)

K(Z) 0.0 1.000 6.0 1.000

'1

12.

0.925 1 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

W4 LF CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F-) (LCO 3.2.2)

F[

shall be limited by the following relationship:

FN <FTP [1.0 + PFv(1.0 - P)]

Where, FP

= FA limit at RATED THERMAL POWER (RTP)

= 1.586 PFw

= power factor multiplier for FA

= 0.3 P

=

THERMAL POWER RATED THERMAL POWER F,

= Measured values of F; obtained by using the movable incore detectors to obtain a power distribution map. The measured values of F, shall be used since an uncertainty of 4% for incore measurement of F; has been included in the above limit.

Page 9 of 15

WuRLF REEK INUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 2.8 Reactor Trip System Overtemperature AT Setpoint Parameter Values (LCO 3.3.1)

Parameter Overtemperature AT reactor trip setpoint Overtemperature AT reactor trip setpoint Tavg coefficient Overtemperature AT reactor trip setpoint pressure coefficient Nominal Tavg at RTP Nominal RCS operating pressure Measured RCS AT lead/lag constant Measured RCS AT lag constant Measured RCS average temperature lead/lag constant Measured RCS average temperature lead/lag constant Value K= 1.10 K2 = 0.0137/ 0F K3 = 0.000671/psig T' _ 586.50F P' > 2235 psig

, =6 sec T2 = 3 sec T3= 2 sec T4 = 16 sec T5 =4 sec T6 =0 sec fl(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

W4LF CREEK FNUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.2)

Parameter Overpower AT reactor trip setpoint Overpower AT reactor trip setpoint Tavg rate/lag coefficient Overpower AT reactor trip setpoint Tavg heatup coefficient Indicated Tavg at RTP (calibration temperature for AT instrumentation)

Measured RCS AT lead/lag constant Measured RCS AT lag constant Measured RCS average temperature lead/lag constant Measured RCS average temperature rate/lag constant Value K4 = 1.10 Ks5= 0.02/°F for increasing Tavg

= 0/'F for decreasing Tavg K6= 0.00128/°F forT> T"

= 0/F for T!<T" T" < 586.50F T, =6 sec T2 = 3 sec T3= 2 sec T6 =0 sec T7 = 10 sec f2(AI) = 0% RTP for all Al Page 11 of 15

4 WLF CREEK

?NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 2.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits (LCO 3.4.1)

Parameter Pressurizer pressure RCS average temperature RCS total flow rate Indicated Value "

Pressure > 2220 psig Tavg -< 590.5 OF 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

W*L,.F CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 APPENDIX A A.

Input relating to LCO 3.2.1:

F0 (Z)m-transient W(Z) =

Ft(Z),,ody,,tate These values are issued in a controlled report which will be provided on request.

Input relating to SR 3.2.1.2 Cycle Burnup (MWD/MTU)

_>0 Cycle Burnup (MWD/MTU)

>0 F.j (Z) Penalty Factor

(%)

2.00 Fw(Z) Exclusion Zone

(%)

Top Bottom 15 15 Page. 13 of 15

W LF CREEK FNUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Revision 0 B.

Approved Analytical Methods for Determining Core Operating 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.

V

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, forthe "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-10216-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).

Page 14 of 15

Wolf Creek Generating Station W*I-,LF CREEK Core e

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'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revisio n 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-1 0266 "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-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. 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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