Text

Submittal of Cycle 24 Core Operating Limits Reports
	ML19302D559
Person / Time
Site:	Wolf Creek
Issue date:	10/22/2019
From:	Benham R; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA 19-0097
	Download: ML19302D559 (37)
	v • d • e

W@LFCREEK

'NUCLEAR OPERATING CORPORATION Ron Benham Manager Nuclear and Regulatory Affairs U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 October 22, 2019 RA 19-0097

Subject:

Docket No. 50-482:

Wolf Creek Generating Station Cycle 24 Core Operating Limits Reports To Whom It May Concern:

These documents are being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station Technical Specifications.

Enclosure I is Revision O of the Wolf Creek Generating Station Cycle 24 Mode 6 Core Operating Limits Report. The Mode 6 only Core Operating Limits Report was issued due to an emergent redesign.

Enclosure II is Revision 1 of the Wolf Creek Generating Station Cycle 24 Core Operating Limits Report applicable to all modes.

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

Sincerely, /J ef!::~

Ron Benham RB/rlt Enclosure I -

Wolf Creek Generating Station Cycle 24 Mode 6 Core Operating Limits Report, Revision 0 Enclosure II - Wolf Creek Generating Station Cycle 24 Core Operating Limits Report, Revision 1 cc:

S. A. Morris (NRC), w/e N. O'Keefe (NRC), w/e B. K. Singal (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/HCNET

Enclosure I to RA 19-0097 ENCLOSURE I WOLF CREEK GENERATING STATION CYCLE 24 MODE 6 CORE OPERATING LIMITS REPORT, REVISION 0 (17 pages)

APF 05-013-01, REV. 04 TR-94-0015 WCNOC Cycle 24 MODE 6 Core Operating Limits Report (COLR)

Revision 0 ENGINEERING REVIEW:

DRAFTER: N/A CHECKER: N/ A ENGINEER: See attached.

SUPERVISOR:1 * ~ ~

10/14/19 0 ELECTRONIC APPROVAL 1.0 APPROVED-MFG. MAY PROCEED 2.0 NOT APPROVED-RESUBMIT FINAL DOCUMENT/DRAWING-MFG. MAY PROCEED 0 YES 0 NO 3.0 APPROVED INFORMATION NOT CONTROLLED UNDER DESIGN PROCESS 4.0 ACCEPTABLE-MAINTAIN AS RECORD (INFO. ONLY)

5. 0 RESTRICTED FOR WOLF CREEK PLANNING ONLY-MFG. MAY PROCEED 0

YES 0

NO APPROVAL OF THIS DOCUMENT/DRAWING DOES NOT RELIEVE SUPPLIER/CONTRACTOR FROM FULL COMPLIANCE WITH CONTRACT, SPECIFICATIONS AND/OR PURCHASE ORDER REQUIREMENTS.

COMMENTS:

VETIP (Al OSC-001 ): This document does not contain design information that requires an engineering Change Package.

Safety Related NOTE: DO NOT RELEASE this document until directed by Nuclear Engineering.

This document is to be released during Refuel 23 after core offload and before core reload.

P.O.#: N/A VENDOR MANUAL:

PAGE: N/A CHANGE PACKAGE#:

INCORPORATED CHANGE DOCUMENT(S):

N/A N/A REV.#

DC RELEASED:

~FCREEK DigsigDSR 3 0.50 W27 DC65 10/14/2019

'NUCLEAR OPERATING CORPORATION COMPONENT NUMBER(S) N/ A COMPONENT NUMBERS ARE FOR INITIAL (REV. W01) DATA LINKING ONLY. ADDITIONAL COMPONENT LINKS ARE MADE IN DATABASE ONLY.

Wft.FCREEK

'NUCLEAR OPERATING CORPORATIQN Wolf Creek Generating Station Cycle 24 Core Operating limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE24 MODE 6 CORE OPERATING LIMITS REPORT Revision 0 October 2019 Prepared by:

10/14/2019 Ian Miller Date Reviewed by:,

I 0 f - 2..., i '1

-+----+-------------

Date Approved by:

Chad Lisle Date Page 1 of 16

~~-~

~LFCREEK

'NUCLEAR OPERATING CORPORATION 1.0 CORE OPERATING LIMITS REPORT Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0 The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 24 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 (SOM) 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)) (Fa Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (Ji!)

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 Page 2 of 16

~FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 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)

I DES 1 and 2, the combination of THERMAL POWER, Reactor Coolant em (RCS) highest loop average temperature, and pressurizer pressure shall ed the limits in Figure 2.1.

Unacceptable Consequences

~ 640 o-i 0

1--

G)

U)

~ 620 600 1925 psia Acceptable Consequences 530-1--...1.-...1.-...1........,........L........L........l...--,---L..--L......J..--,-....J..----l..___...___,.___.___.~.___,---.,__.,__-'---I 0.

0.2 0.4

~6

~8 F roction of Roted Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16 1.2

W$LFCREEK Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0

'NUCLEAR OPERATING CORPORATION 2.2 Moderator Temperature Coefficient (MTC) (LCO 3.1.3, SR 3.1.3.1, 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 rawn, RATED THERMAL POWER condition).

0 60 PM MTC Surveillance limit is -46 pcm/°F (equilibrium, all rods n RATED THERMAL POWER condition).

6.0, 70%

OPERATION 10 20 30 40 50 60 70

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

THERMAL POWER(%)

Page 4 of 16 UNACCEPTABLE OPERATION

MLFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 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)

T Control Bank insertion limits are specified in Figure 2.4. The Control Bank dra al sequence is A-B-C-D. The insertion sequence is the reverse of the d

al sequence. The difference between each sequential Control Bank s

T 200 180 E 14 0 p

s W 12 0 I

T H 100 D

R A

w N

80 60 40 20 0

steps when not fully inserted and not fully withdrawn. -

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t 3 Q.12 % : Q ) I 0

2 0 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<!: 222 and :S: 231 steps withdrawn.

Page 5 of 16

\\N$1.FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 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.

0 F

R A

T E

90 D 80 T

H E

R 70 M

A L

p 60 0

w E

R 50 40

-40

( -2 9, 5 0 )

-30

( 5, 10 0 )

UNACCEPTABLE OPERATION

-20

-10 0

10 20 AXIAL FLUX DIFFERENCE (%~I)

Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of Tl'IERMAL POWER(%)

Page 6 of 16 30 40

W$1.FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor {Fo(Z))(F9 Methodology) (LCO 3.2.1, SR 3.2.1.1, SR 3.2.1.2)

F0 (Z) ~ CFQ *K(Z), for P > 0.5 p

J ~ c;; *K(Z), for P ~ 0.5 K(Z)

Ft" (Z) is the measured value o measurement obtained with the Power Distribution Monitoring System Manufacturing tolerances are accounted for in the 1.

factor. Measurement uncertainty for MIDS is accounte r

PDMS measurement uncertainty is accounted for in the Ua determined by PDMS.

F/! (Z)=FJ.(ZJW(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, F~v (Z) uses FJ-'(Z) that is determined from an F;1 (Z) that reflects full-power steady-state conditions rather than current conditions.

See Appendix A for: FQ Penalty Factor.

Page 7 of 16

MLFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0 r-0

<(

u.

(!) z 1.2

~

< 0.6 w

D.

C w

~ 0.4

<(

E 0::

0 z 0.2 0.0 0

I I

I

i----------~----------

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--,----------1 ----------T----------T----------,----------

6.0 1.000 12.0 0.925 2

4 10 12 Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Page 8 of 16

W8LFCREEK Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0

'NUCLEAR OPERATING CORPORATION

2. 7 Nuclear Enthalpy Rise Hot Channel Factor ( F:,) (LCO 3.2.2)

F! shall be limited by the following relationship:

FN ~ F,:P[l.O+PF,viI.O-P)]

F! limit at RATED THERMAL POWER (RTP) p

=

When F! is obtaine e measured value is increased by an uncertai factor (U11 and the factor is determined by PDMS, with a Page 9 of 16

W$1.FCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 2.8 Reactor Trip System Overtemperature ~T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 1)

Parameter Ov rtemperature ~ T reactor trip setpoint rtemperature ~ T reactor trip setpoint Tavg C

ffi.

t pressure Measured R

~ T lead/lag constant Measured RCS average constant 0.0184 / %RTP {(qrqb) - 5% RTP}

Where, qt and qb are percent RTP in the upper and lo respectively, and qt+ qb is the total THERMAL P Page 10 of 16 Value K1 = 1.10 K2 = 0.0137/°F KJ = 0.00095/psi T' ~ 586.5°F P' 2:: 2235 psig

-r1 = 6 sec

-r2 = 3 sec

-r3 = 2 sec

-r4= 16 sec

-rs=4 sec

-r5 = O sec

core,

MLFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower~ T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 2)

Parameter f2(~l) = 0% RTP for all ~I Page 11 of 16 Value Ki=1.10 K5 = 0.02/°F for increasing T avg

= 0/°F for decreasing T avg K5= 0.001281°F for T > T"

= 0/°F for T ::; T" T" ::; 586.5°F 1:1 = 6 sec 1:2 = 3 sec 1:3 = 2 sec 1:5 = 0 sec

\\Na.FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 0 2.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling {DNB)

Limits (LCO 3.4.1)

R~ ~13~ tEJnPltiif'LrT'J' [Ia~90.6 °F (Average of 4 channels)

J-\\.1 '-I J-\\. L I LC. I.Jsgo.8 °F (Average of 3 channels)

RCS total flow rate Flow 2 376,000 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)

S1!CTT61Mall2 :r1t2aNt,lf1300 pcm ANALYZED Page 12 of 16

Ml.FCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION APPENDIX A A.

Input relating to LCO 3.2.1:

Core Operating Limits Report Revision 0 Fa (Z)max transient l

W(Z) = -

x-for P > 0.5 FQ (Z)'teadystate p'

Fa (Zr"" transient 1

W(Z)= -

d x-, forP::; 0.5 FQ (zt** ystate 0.5 Page 13 of 16

~LFCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Input relating to SR 3.2.1.2 Cycle Burnup (MWD/MTU)

~,O to ::; 7859 Cycle Burnup (MWD/MTU)

7,000

> 7,000 FQ (Z) Penalty Factor

(%)

2.00 2.23 2.21 2.15 2.06 2.00 15 [11]

10 [7]

Page 14 of 16 Core Operating Limits Report Revision 0

MLFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 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.

WCAP-11397-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-11397, Revised Thermal Design Procedure."

2.

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 - F0 Surveillance Technical Specification" (TAC No. M88206).

3.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

NRC Safety Evaluation Report dated May 28, 1985, "Acceptance for Referencing of Licensing Topical Report WCAP-9272(P)/9273(NP), Westinghouse Reload Safety Evaluation Methodology."

4.

WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," Revision O, 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)."

5.

WCAP-16045-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-16045-P, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON."

6.

WCAP-16045-P-A, Addendum 1-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)."

7.

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."

Page 15 of 16

W$1.FCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0

8.

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 ReportWCAP-12610, Appendix B, Addendum 1, 'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO.

M86416)."

9.

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZirloTM," July 2006.

NRC Safety Evaluation dated June 10, 2005, "Final Safety Evaluation for Addendum

-1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A, "Optimized Zirlo',"

(TAC NO. MB8041)."

10. WCAP-87 45-P-A, "Design Bases for the Thermal Overpower I',. T and Thermal Overtemperature 1',.T 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 1',.T and Thermal Overtemperature 1',.T Trip Functions."'

Page 16 of 16

Enclosure II to RA 19-0097 ENCLOSURE 11 WOLF CREEK GENERATING STATION CYCLE 24 CORE OPERATING LIMITS REPORT, REVISION 1 (17 pages)

APF 05~013~01, REV. 04 TR-94-0015 WCNOC Cycle 24 Core *Operating Limits.Report (COLR)

Revision 1 ENGINEERING REVIEW:

DRAFTER: N/ A CHECKER: N/ A ENGINEER: See attached.

SUPERVISOR: ~

~

10/19/2019 0 ELECTRONIC APPROVAL 1.0 APPROVED-MFG. MAY PROCEED 2.0 NOT APPROVED-RESUBMIT FINAL DOCUMENT/DRAWING-MFG. MAY PROCEED 0

YES 0

NO 3.0 APPROVED INFORMATION NOT CONTROLLED UNDER DESIGN PROCESS

4. 0 ACCEPTABLE-MAINTAIN AS RECORD (INFO.ONLY)

5. 0 RESTRICTED FOR WOLF CREEK PLANNING ONLY-MFG. MAY PROCEED 0

YES 0

NO APPROVAL OF THIS DOCUMENT/DRAWING DOES NOT RELIEVE SUPPLIER/CONTRACTOR FROM FULL COMPLIANCE WITH CONTRACT, SPECIFICATIONS AND/OR PURCHASE ORDER REQUIREMENTS.

COMMENTS:

VETIP (Al 05C-001 ): This document does not contain design information that requires an engineering Change Package.

Safety Related NOTE: DO NOT RELEASE this document until directed by Nuclear Engineering.

This document is to be released durinQ Refuel 23 after core offload and before core reload.

P.O.#: N/A VENDOR MANUAL:

PAGE: N/A CHANGE PACKAGE#:

INCORPORATED CHANGE DOCUMENT(S):

N/A N/A REV.#

DC RELEASED:

~FCREEK DigsigDSR 3 0.50 W28

'NUCLEAR OPERATING CORPORATION COMPONENT NUMBER(S) N/ A COMPONENT NUMBERS ARE FOR INITIAL (REV, W01) DATA LINKING ONLY. ADDITIONAL COMPONENT LINKS ARE MADE IN DATABASE ONLY.

W@JLFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 WOLF CREEK GENERATING STATION CYCLE 24 CORE OPERATING LIMITS REPORT Revision 1 October 2019 Prepared by:

10/14/2019 Ian Miller Date Date Approved by:

la/12/2.oir Chad Lisle Date Page 1 of 16

WM.FCREEK

'NUCLEAR OPERATING CORPORATION 1.0 CORE OPERATING LIMITS REPORT Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 24 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 (SOM) 3.1.3 Moderator Temperature Coefficient (MTG) 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)) (Fa Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F!)

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 Page 2 of 16

-LFCREEK Wolf Creek Generating Station Cycle 24

Core Operating Limits Report Revision 1

'NUCLEAR OPERATING CORPORATION 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

1-640 0"1 0

f---

Q)

U)

U) 5 620 600 1925 psia 2460 psia Unacceptable Consequences Acceptable Consequences 580-+-......... --------.-................... __.____,..__.___.____.___,..___.___.____..~____.____.____.--,-__,..............................j 0

0.2 0.4 0.6 0.8 F roction of Roted Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16 1.2

~

W$LFCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 1 2.2 Moderator Temperature Coefficient (MTC) (LCO 3.1.3, SR 3.1.3.1, 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 ii:"

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4

, 5 Ill

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0 5 2 Ill CII 0

E 0

0 10 20 I

I UNACCEPTABLE OPERATION 6.0, 70%

ACCEPTABLE OPERATION 30 40 50 60 70 80 90

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

THERMAL POWER(%)

Page 4 of 16 100

-LFCREEK Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1

'NUCLEAR OPERATING CORPORATION 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 limits are specified in Figure 2.4. The Control Bank withdrawal sequence is A-B-C-0. The insertion sequence is the reverse of the withdrawal sequence. The difference between each sequential Control Bank position is 115 steps when not fully inserted and not fully withdrawn.

(FULLY WITHDRAWN) 220 200 180 160 s

T E 140

p s

w 120 I

T H 1 0 0 D

R A

80 w

N 60 40 20 0

0

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__.L __..l __..J ___ I_ __ L L __ J. __.J __ -' ___ 1 ___ L __ L __..l __..J __ _ L __ L __..L __.J __

1 I

I l

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I

--L--..l--~---1--

L--L--..L--~---1---1---L--L--..l--..J--

1---L--L--..L--~---

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I I

I J

I I

I 1

I I

I J

--~--..1--~--

---~--~--+--~--~---J---L--~--..1-- ---J---~--~--~--~--

1 I

I I

I 1

1 1

t 1

1 1

B1AN1K I

r 1

1

~ -

I -

~ -

- i -- ~ --

I - -~ -

~ -

--+-- --~---1---~--~--+--~--4---1---~-- --~--~---,---~--~--~--~--

I t

I I

I "t" -

-r -

I -

T -

I -

I I

I

- r - -,. - - -, - - -

- r - - r - -

T -

- - - r - - r - -

1 -

- r - - r - -

T -

Qo/~, 4~)

I 1

I I

I

- - r - - -, -

- r - - r - - T -

-,- - - r - - r - - -, -

- r - - r - - T -

I I

I l

l I

I I

--r--,--~---i---r--r--T--~- ~---i---r--r--,---,---,---r--r--T--~--

I I

I 1

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I l

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I I -

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1 J

l I

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-- t --,-- t---1---1 --, -- 1 -- 1-- 1------1 --1 -- 1-- 1------1 --1 --

I -- 1---

1 3 Q.'2 %

I Q ) I I

[

I 20 40 60 80 100 (FULLY INSERTED)

THERMAL POWER (Pe re en t)

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 H>

Wft.F CREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 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 UNACCEPTABLE 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

( -2 9, 5 0 )

40

-40

-30

( -1 5, 100 )

( 5, 100 )

UNACCEPTABLE OPERATION ACCEPTABLE OPERATION

( 24, 50 )

-20

-10 0

10 20 AXIAL FLUX DIFFERENCE(% Lil)

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

Page 6 of 16 30 40

\\N$LFCREEK 1NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 2.6 Heat Flux Hot Channel Factor (Fo(Z)}(Fa Methodology) (LCO 3.2.1, SR 3.2.1.1, SR 3.2.1.2)

F0 (Z) ~ CFQ *K(Z), for P > 0.5 p

FQ(Z) ~ c;.; *K(Z), for P ~ 0.5 where, P THERMAL POWER

=

RATED THERMAL POWER CFQ

=

FQ(Z) limit at RATED THERMAL POWER (RTP)

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

Ft1 (Z) is the measured value of FQ(Z), inferred from a power distribution measurement obtained with the Movable lncore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS).

Measurement uncertainty is applied as follows.

Ff(Z)=FQM(Z)(I.03)(1.05)=Ft1(Z)(I.08I5) when Ft1(Z) is obtained from MIDS.

Ff(Z)=Ft1(Z)(I.03)(UQu) when Ft1(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 Uou factor, and it is determined by PDMS.

F! (Z)=Ff(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, F! (Z) uses Ff (Z) that is determined from an Ft1 (Z) that reflects full-power steady-state conditions rather than current conditions.

See Appendix A for: FQ Penalty Factor.

Page 7 of 16

\\NeLFCREEK Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 N

st 0:::

0

~

u

~

C) z 2 <

w 0..

C w

N :J 0:::

0 z

'NUCLEAR OPERATING CORPORATION 1.2 1.0

i----------~----------

1 0.8 I

I l

I I

r----------r----------T----------T----------1----------

I 0.6

~----------~----------+----------+----------4----------

1 I

0.4 I

I I

- -I---------- --------

FQRTP = 2.50 0.2

~ - - - - - - - - - -

+- - - - - - - - - - -i ----------i- - Elevation (ft)

K(Z)

I 0.0 1.000 6.0 1.000 12.0 0.925 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 16

weLFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1

2. 7 Nuclear Enthalpy Rise Hot Channel Factor ( F~) (LCO 3.2.2)

F; shall be limited by the following relationship:

F; ~ F,:;t[1.o+PF,vA1.o-P)]

Where, F/:;1/

= F; limit at RATED THERMAL POWER (RTP)

= 1.650 PF Ml = power factor multiplier for F; p

= 0.3

=

THERMAL POWER RATED THERMAL POWER F; is the measured value of F;, inferred from a power distribution measurement obtained with the Movable lncore 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 F; is obtained from PDMS, the measured value is increased by an uncertainty factor (U~H), and the factor is determined by PDMS, with a lower limit of 4%.

Page 9 of 16

\\N$LFCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 1 2.8 Reactor Trip System Overtemperature ~ T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 1)

Parameter Overtemperature ~ T reactor trip setpoint Overtemperature ~T reactor trip setpoint Tavg coefficient Overtemperature ~ T reactor trip setpoint pressure coefficient Nominal T avg (Tref from Rod Control) at RTP Nominal RCS operating pressure Measured RCS ~ T lead/lag constant Measured RCS ~ T lag constant Measured RCS average temperature lead/lag constant Measured RCS averag~ temperature lead/lag constant Value K1=1.10 Kz = 0.0137/°F K3 = 0.00095/psi T' ~ 586.5°F P' ~ 2235 psig

't1 = 6 sec

't2 = 3 sec

't3 = 2 sec

't4= 16 sec

't5 = 4 sec

't6 = 0 sec f1(~I) = -0.0227 / %RTP {23% RTP + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP ~ (qt-qb) ~ 5% RTP 0.0184 / %RTP {(qt-qb) - 5% RTP}

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

\\N$LFCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 1 2.9 Reactor Trip System Overpower~ T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 2)

Parameter Overpower~ T reactor trip setpoint Overpower ~T reactor trip setpoint Tavg rate/lag coefficient Overpower~ T reactor trip setpoint T avg heatup coefficient Nominal Tavg (Tref from Rod Control) at RTP Measured RCS ~ T lead/lag constant Measured RCS ~ T lag constant Measured RCS average temperature lag constant Measured RCS average temperature rate/lag constant fa(~I) = 0% RTP for all ~I Page 11 of 16 Value Ks= 0.02/°F for increasing Tavg

= 0/°F for decreasing T avg K6 = 0.00128/°F for T > T"

= 0/°F for T ~ T" T" ~ 586.5°F

-c1 = 6 sec

-cz= 3 sec

-c3= 2 sec

-c5= 0 sec

Ml.FCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 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 ~ 2219 psig (Average of 4 channels)

~* 2221 psig (Average of 3 channels)

RCS average temperature Tavg ~ 590.8 °F (Average of 4 channels)

~ 590.6 °F (Average of 3 channels)

RCS total flow rate Flow~ 376,000 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% ~k/k).

Page 12 of 16

~LFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 APPENDIX A A.

Input relating to LCO 3.2.1:

Fa (Z)max transient l

W(Z) = -

x-fior P > 0.5 FQ (Zyteadystate p,

_ Fa (Z)max transient l

W(Z) = -

d x-, fior P ::; 0.5 FQ (Ztea ys1a1e 0.5

where, P= RATED THERMAL POWER THERMAL POWER FQ(Z)maxtransie/11 = Maximum (FQ(Z) X p) calculated over the entire range of power shapes analyzed for Condition I operations (p = power at which maximum occurs).

FQ(Z)"eadystate = (FQ(Z) x p) 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 Fo(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

\\M&LFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 Input relating to SR 3.2.1.2 Cycle Burnup FQ (Z) Penalty Factor (MWD/MTU) 2: 0 to :5 7859 8057 8254 8452 8650 2: 8847 Cycle Burnup (MWD/MTU)

5 7,000

> 7,000

(%)

2.00 2.23 2.21 2.15 2.06 2.00 FQ (Z) Exclusion Zone

(% [INCORE mesh points])

Top Bottom 15 [11]

10 [7]

Page 14 of 16 15 [11]

10 [7]

\\N$LFCREEK

'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 24 Core Operating Limits Report Revision 1 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.

WCAP-11397-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-11397, Revised Thermal Design Procedure."

2.

WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control - Fa 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).

3.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

NRC Safety Evaluation Report dated May 28, 1985, "Acceptance for Referencing of Licensing Topical Report WCAP-9272(P)/9273(NP), Westinghouse Reload Safety Evaluation Methodology."

4.

WCAP-16009-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)."

5.

WCAP-16045-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-16045-P, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON."

6.

WCAP-16045-P-A, Addendum 1-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)."

7.

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."

Page 15 of 16

Wft.FCREEK Wolf Creek Generating Station Cycle 24

'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 1

8.

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)."

9.

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized Zirlo'," July 2006.

NRC Safety Evaluation dated June 10, 2005, "Final Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A, "Optimized Zirlo',"

(TAC NO. MB8041)."

10. WCAP-87 45-P-A, "Design Bases for the Thermal Overpower~ T and Thermal Overtemperature ~ T 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 d T and Thermal Overtemperature d T Trip Functions."'

Page 16 of 16

ML19302D559

Text

Subject:

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