ET 05-0005, Cycle 15 Core Operating Limits Report

Cycle 15 Core Operating Limits Report

ML051370300
Person / Time
Site:	Wolf Creek
Issue date:	05/05/2005
From:	Garrett T Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 05-0005
Download: ML051370300 (16)

Text

- WOLF CREEK 'NUCLEAR OPERATING CORPORATION Terry J. Garrett Vice President Engineering May 5, 2005 ET 05-0005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 15 Core Operating Limits Report Gentlemen:

Enclosed is Revision 0 of the Wolf Creek Generating Station Cycle 15 Core Operating Limits Report (COLR). This document is being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station Technical Specifications.

There are no commitments associated with this submittal. Please contact me at (620) 364-4084 or Mr. Kevin Moles at (620) 364-4126 for any questions you may have regarding this submittal.

Very truJours, ferry J. Garrett TJG/rIg Enclosure cc: J. N. Donohew (NRC), w/e D. N. Graves (NRC), w/e B. S. Mallett (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411/ Burlington, KS 66839 Phone: (620) 364-8831 An Equal Opportunity Employer MIF/HCIVET

Wolf Creek Generating Station W&LF CREEK

'NUCLEAR OPERAlING CORPORATION Cycle 15 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 15 CORE OPERATING LIMITS REPORT Revision 0 April, 2005 Prepared by:

304 / - -------

C CL U'4/12uzuu Matthew K. Morris Date Reviewed by: 04/05/2005 William Kennamore Date Approved by: 47' ' 04/05/2005 D n MI -------

• _+

rs. Li. I-ldIlII lqdl I UdLe Page 1 of 15

Wolf Creek Generating Station W~LF CREEK NUCLEAR OPERATING CORPORATON CreOprain Liit Core Operating Limits Repo0t Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 15 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 Alignments 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 PHYSICS TEST Exceptions - MODE 2 3.2.1 Heat Flux Hot Channel Factor (FQ(z)) (FQ Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (EH) 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.LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 15 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 I 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 C

oI V 620

.2

.0 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fractlon of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 15

Wolf Creek Generating Station WELF CREEK'NUCLEAR OPERATING CORPORATION Cycle 15 Core Operating Limits Report Revision 0 2.2 Moderator Temperature Coefficient (MTCQ (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/OF.

The 300 PPM MTC Surveillance limit is -41 pcm/OF (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 I I I.UNACCEPTABLE OPERATION I I 6.0. 70%

M.

0 a2 0.

ACCEPTABLE OPERATION Mu 0:

',' ._3I 0 10 20 30 40 50 60 70 80 90 100

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

RATED THERMAL POWER Page 4 of 15

Wolf Creek Generating Station W'LF& NUCLEAR OPERATING CORPORATION CREEK Cycle 15

-Core Operating Limits Report Revision 0 2.3 Shutdown Bank Insertion Limits (LCO 3..1.5)

The shutdown banks shall be fully withdr awn (i.e., positioned within the interval of

> 222 and < 231 steps withdrawn).

2.4 Control Bank Insertion Limits (LCO 3.1.( 3)

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

(FULLY WITHDRAWN)

II I I I /(76~7%I.222) I i I I I I I I , ' I .

_____Lls_____l__

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 2-- _-- ,-- -- --

I I I, , , .,I I IL /.__ __J__I__I _L__IJ AI A I I - I I I I I I I I I I I I .

80 B I ' ( 100%'. 161)

TI T *I I I I I I I I ,. . . * . .I * . . .

1 I I ,

I ,

I IL - I L'

,- - -l I - '- Js - I -

I E 140 ._ ' I L--I I ,~ , , , . , , , I I I I L__JJ__ 1 - _)l 1- JB .

WI I I I I I I I I S ... L__l__J l__L__.J_

I . I. . * . . . .. I

• I I I I I I I I I , ,. , , , ,I , I T I PI I.

, I I - I I I I I I H 140 I - I -J- I- -- I I L__.._JI I IC

, , . I . Is I oF

+/-.46..

l

• L.L.I..J

. I I I

_Ll__________LIJ_

W 80 ___ __._____-. I I I D Il I I I

_I I I I I I I ,I ,I I I I IX Ia I I I I I I I I I I I I I

_____ __Jll___l__

I 620 _ .1 - -L LI-I -J-I I I I I I o I I I YI I I I I I I I I I I , , , , ., , , I A 0 - 20i-' L+/-40..

(FULLY INETD I ~~ ~ RAEI 1 I _

TEMI

_ I I I I .1 1 1 1 1 I I , ., , I, , I 20 ___JJ______L__1__J_ 0L__L_ _ -. J- _____Ll____e__L__

I , , , ., , , , , I

1) ILI I 30§ I I 0I II I I _I I I I I I I I I _J_ IL__L l ,. ,

I , . , . . . . . I 60 80 100 L 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 2 222 and S 231 steps withdrawn.

Page 5 of 15

Wolf Creek Generating Station WOLF CREEK rNUCIR OPERATING CORPORATION Cycle 15 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.

110 100 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

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

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

Wolf Creek Generating Station W&LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 15

.Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (FdZ))(Fo Methodologv) (LCO 3.2.1, SR 3.2.1.2)

FQ(Z) < C *K(Z), forP> 0.5 P

FQ (Z)S - *K(Z), forP < 0.5 0.5 THERMAL POWER RATED THERMAL POWER.

CFQ =FR m FQR = FQ(Z) limit at RATED THERMAL POWER (RTP)

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

.Fc(Z) = F"'(Z)(1.03)(1.05) =FQ(Z)(1.08 15) where, F,'(z)= Measured value of F,/Z) from incore flux map FQw Z) FQC(Z)WaZ) where, RWZ) = 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

Wolf Creek Generating Station W',kF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 15 Core Operating Limits Report Revision 0 1.2 I I I I . I I I .I N 1.0 i- I I I 0.8 ----------- L-----------------------

I. I CD Z I1 I I z 0 .6 ILI -- - - - - ,. - - - .- --.I -- - - -

Lu .I I I N .0.4 ------------

.I r---_____ -I _ ___

---

I 0 I . . I I I . I

. I I

° 0.2 -------------------------------

I I 0.2

0. I . . 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 15

Wolf Creek Generating Station W4LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 15 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpv Rise Hot Channel Factor (Fa) (LCO 3.2.2)

F.', shall be limited by the following relationship:

F.,N, < F"[T1.0 + PFA,(1.0 - P)]

Where, FRp = F.,,, limit at RATED THERMAL POWER (RTP)

= 1.586 PFw = power factor multiplier for F,.H

= 0.3 P = THERMAL P0 IER RATED THERMAL POKER F. Measured values of FAH 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

Wolf Creek Generating Station

' L Cycle 15

'NUCLEARl OPERATING CORPORATION Core Operating Limits Report Revision 0 2.8 Reactor Trip System Overtemnerature AT Setpoint Parameter Values (LCO 3.3.1)

Parameter Value Overtemperature AT reactor trip setpoint K, = 1.10 Overtemperature AT reactor trip setpoint Tavg K2 = 0.0137/OF coefficient Overtemperature AT reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal Tavg at RTP T' < 586.50F Nominal RCS operating pressure P' 22235 psig Measured RCS AT lead/lag constant ¶i =6 sec 2=3 sec Measured RCSAT lag constant T3 = 2 sec Measured RCS average temperature lead/lag ¶4= 16 sec constant ¶5=4 sec Measured RCS average temperature lead/lag ¶6=0 sec constant f1(Al) = -0.0227 {23% + (qt-qb)} when (qrqb) < -23% RTP 0%of RTP when -23% RTPS* (qt-qb) S 5% RTP 0.0184 {(qrqb) - 5%} when (qrqb)> 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

Wolf Creek Generating Station W4LF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 15 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 = 0.02/0F for increasing Tavg rateAag coefficient = O/F for decreasing Tavg Overpower AT reactor trip setpoint T8v, heatup Kr,= 0.001280F for T> T" coefficient -

0 for T

• T"

/F Indicated Tavg at RTP (calibration temperature rf < 586.50F for AT instrumentation)

Measured RCS AT lead/ag 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/ag T7= 10 sec constant f2(AI) = 0%RTP for all Al*

Page 11 of 15

.I Wolf Creek Generating Station VWELF CREEK

'NUCLEAR OPERATING cORPnoRAON Cycle 15 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 2 2220 psig RCS average temperature Tavg<590.5OF RCS total flow rate Flow 2 371,000 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*

WZLF CREEK

'NUCLEAR OPERATING CORPORATION Cycle 15 Core Operating Limits Report Revision 0 APPENDIX A A. Input relating to LCO 3.2.1:

t F (Z)

F (Z) teadySUte These values are issued in a controlled report which will be provided on request..

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

0 to 20574 2.00 Note: All cycle burnups outside of the above table shall use a 2% penalty factor for compliance with SR 3.2.1.2. Linear interpolation should be used for intermediate cycle burnups.

Page 13 of 15

Wolf Creek Generating Station NUCLFAR OPERATING CORPORATION 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.

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."

4. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control - F0 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 - Fa 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 A Fte WIULF "CREE

'UCLEAR SA K

OPERATING CORPORTONRevision Cycle 15 Core Operating Limits Report 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-10266, 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 CodeAddendum 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, "Qual ification 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 Corn puter 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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