Text

Submittal of Cycle 22 and Cycle 23 Core Operating Limits Report
	ML18127A072
Person / Time
Site:	Wolf Creek
Issue date:	04/29/2018
From:	Hafenstine C; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA 18-0054
	Download: ML18127A072 (35)
	v • d • e

NUCLEAR OPERATING CORPORATION Cynthia R. Hafenstine Manager Nuclear and Regulatory Affairs April 29, 2018 RA 18-0054 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington , DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 22 and Cycle 23 Core Operating Limits Report To Whom It May Concern :

Enclosure I is Revision O of the Wolf Creek Generating Station (WCGS) Cycle 23 Core Operating Limits Report (COLR) . This document is being submitted pursuant to Section 5.6.5 of the WCGS Technical Specifications.

Enclosure II is Revision O of WCGS Cycle 22 COLR. During preparation of Cycle 23 WCGS COLR, it was identified that the Cycle 22 WCGS COLR was not submitted pursuant to Section 5.6.5 of the WCGS Techn ical Specifications. This has been captured in the Corrective Action Program .

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

Sincerely, trn}ftlL re JJ{fw11 Cynth ia R. Hafenstine

)v_,

CRH/rlt Enclosure I - WCGS Cycle 23 Core Operating Limits Report Enclosure II - WCGS Cycle 22 Core Operating Limits Report cc: K. M. Kennedy (NRC), w/e B. K. Singal (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/HC/VET

Enclosure I to RA 18-0054 ENCLOSURE I WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT Revision 0 April 2018 Prepared by: ~ ~ 4/24/2018 Ian Miller Date Reviewed by: /7~

Dustin Wi rth

~ 04/24/2018 Date Approved by: ~1~ 04/24/2018 Gregory S. Kinn Date Page 1 of 16

Wolf Creek Generating Station W$LFCREEK 1

NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 23 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 (F0 (z)) (F a 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 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

Wolf Creek Generating Station W$ LFCREEK 'NUCLEAR OPERATING CORPORATION Cycle 23 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 Unaccept a bl e Operation 660 - - --

- -' - 2400 p s ia

~------I'-- '

r----...__

- ' -~

/

6 40 -........_ -' '

2 000 p s ia U:::-

':!..- / ------- ~ ;--...,.

........__ r---_

~

Cl

~

I---""'" - '*

i-:'

.,,.,, r--.. 225 0 p s ia

~ ' *.

Q)

Q) 620 I'--..

~ ~ ------ ~

\

Q) 1925 p s ia

I'-- \.

' ' \. .

0 cu

3: r-----..

..2

<t: ----r----.. ---- ~

~

\

\.

600

--r--._ \. '

Acceptable Operation '\\

~

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

Wolf Creek Generating Station W lLFCREEK 'NUCLEAR OPERATING CORPORATION Cycle 23 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 pcml°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 UII ACCEPTJ BLE DPERATI< ~N E 6 .0 , 70%

u C.

- 6 z

u ii:

II.

w 0

u w

a:

, 4 1-4 a: A<tCEPTAB E w

a. CPERATIO~

w l-a:

0 1-4 2 a:

w Q

0

ii:

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

% of RA TED T HERM AL POWER Figure 2.2 Moderator Temperature Coefficient Vs.

THERMAL POWER(%)

Page 4 of 16

Wolf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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) 2 20 *( 2 1 . 7 uA , 2 tL ) *( 7 1 .7 0/c . 2, 2) i,

/ ~

200 / /

V V

,/

~A ....... /

180

,/v a ~v V V 160

,t7

( o ob , 1 6 1 )

f-V ( 1 p o °/c

1E 1 )

s T

,- I/ I---- V E 140 p

/e AN Ii.< /

/ C s

W 120 V

/v V

>~/

T H 100

,V

/ t--

~

V t----

D ,__ V ,-

V V

~ 1-- ,-

R A

w 80 l/E AN K L

[

N - / *- - - - - - ,_ ,-. ,_ - ,-

60 / ,/

/ /

V 1(0 ~ >. 4 3 )

40 /

V 20 V - ,- -

- - V - - ~- - - -

0 30 2 °/o 0) [7 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 w ithdrawn shall be the condition where control banks are at a position within the interval of ~ 222 and ~ 231 steps withdrawn.

Page 5 of 16

Wolf Creek Generating Station Wel.FCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 2.5 AXIAL FLUX DIFFERENCE (AFD) (Re laxed 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 O/o 0

F 90 R

A T

E D80 T

ACCEPTABLE H

OPERATION E

R70 M

A L

p 60 0

w E

R 50

( -2 9 , 50) ( 24 , 50) 40

-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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (F0 (Z))(Fo Methodology) (LCO 3.2.1, SR 3.2.1.2)

F0 (Z) :s; CFQ *K(Z), for P > 0.5

- p FQ(Z) :s; c;; *K(Z), for P :s; 0.5 THERMAL POWER where , P =

RA TED THERMAL POWER CFQ = FQ RTP FQ R7P

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

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

FQM (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 .

Measurement uncertainty is applied as follows .

Ff (Z)=F;1 (Z)(l.03)( 1.05)=F;1(Z)(1.0815) when FQM (Z) is obtained from MIDS.

Fi *(Z) = F;1 (Z)(l.03)(UQu ) when F;1 (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 Uau factor, and it is determined by PDMS.

F~v(Z)=Fi °(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/f (Z) uses Ff (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 Wft.FCREEK'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 1.2 N 1.0 S2' 0:::

0

(.)

<(

0.8 ~ - - ------- - - - - - - - '- - - - - - -

LL c.,

z

si::: '

<( 0.6 ,_ ----------

w a.

Q w

N 0.4 ~ -- -

i

<(

~

0:::

0 z 0 .2 ~ Bevation (ft) K(Z) 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

Wolf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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 :

N FM{~ FM,R71' [1.0 + PFM{ ( 1.0- P) ]

Where , Fdi1' = F; limit at RATED THERMAL POWER (RTP)

= 1.650 PFM{ = power factor multiplier for F;

= 0.3 p = 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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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 Value Overtemperature ~ T reactor trip setpo int K1=1.1 0 Overtemperature ~ T reactor trip setpoint T avg K 2 = 0.0137/°F coefficient Overtemperature ~T reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal T avg at RTP T' ~ 586 .5°F Nominal RCS operating pressure P' ~ 2235 psig Measured RCS ~T lead/lag constant -c1 = 6 sec

-c2 = 3 sec Measured RCS ~T lag constant -c3 = 2 sec Measured RCS average temperature lead/lag *4 = 16 sec constant -cs= 4 sec Measured RCS average temperature lead/lag *6 = 0 sec constant 0% of RTP when -23% RTP ~ (q1-qb) ~ 5% RTP Where , q 1 and qb are percent RTP in the upper and lower halves of the core ,

respectively, and q 1 + qb is the total THERMAL POWER in percent RTP.

Page 10 of 16

Wolf Creek Generating Station

\NeLFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower Ll T Setpoint Parameter Values (LCO 3.3.1 ,

Table 3.3.1-1 , Note 2)

Parameter Value Overpower Ll T reactor trip setpoint K4 = 1.10 Overpower Ll T reactor trip setpoint T avg K 5 = 0.02/°F for increasing T avg rate/lag coefficient = 0/°F for decreasing T avg Overpower T reactor trip setpoint T avg heatup K 6 = 0.00128/°F for T > T" coefficient = 0/°F for T ~ T" Indicated T avg at RTP (calibration temperature T" ~ 586 .5°F for Ll T instrumentation)

Measu red RCS LlT lead/lag constant *1 = 6 sec

2 = 3 sec Measured RCS LlT lag constant *3 = 2 sec Measured RCS average temperature lead/lag *6 = 0 sec constant Measured RCS average temperature rate/lag *7 = 10 sec constant f 2{Lll) = 0% RTP for all Lll Page 11 of 16

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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 T avg ~ 590.5 °F 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% Lik/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 16

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 APPENDIX A A. Input relating to LCO 3.2.1:

Fa (Z) max transient }

W(Z)= - x- forP > 0.5 Fg (Z)51eadystate p, F (Z) max transient I 0

W(Z)- d x _ , forP :S 0.5 FQ(Zftea ystate 05 THERMAL POWER where , P =

RATED T HERMAL POWER F o(Z)"" = Maximum (F Q(Z) x p) calculated over the entire range of power shapes

- lransienr analyzed for Condition I operations (p = power at which maximum occurs) .

F g (Zfeady slale = (F Q(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, rathe r than the full-power W(z) values. For these part-power W(z) va lues , the F Q(zf teady 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$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 Input relating to SR 3.2.1 .2 Cycle Burnup FQ(Z) Penalty Factor (MWD/MTU) (%)

~ 0 to :5 7658 2.00 7856 2.14 8053 2.37 8251 2.63 8449 2.86 8646 2.79 8844 2.57 9041 2.32 9239 2.06

~ 9437 2.00 FQ(Z) Exclusion Zone

(% [INCORE mesh points])

Cycle Burnup (MWD/MTU) Top Bottom

5 8,000 15 [11] 15 [11]

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

Page 14 of 16

Wo lf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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-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."

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-10216-P-A , Revision 1A, "Relaxation of Constant Axial Offset Control - F 0 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 15 of 16

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0

7. 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). "

8. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON ," August 2004.

NRC Safety Evaluation dated Ma rch 18, 2004, "Final Safety Evaluation for Westinghouse Topical Report WCAP-16045-P, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON ."

9. 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 fo r Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16045-P-A, Addendum 1, "Qual ification 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," Apri l 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.

M864 16) ."

12. 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, "Optim ized Zirlo',"

(TAC NO. MB8041) ."

13. WCAP-87 45-P-A, "Design Bases for the Thermal Overpower 6. T and Thermal Overtemperature 6. T Trip Function ." September 1986.

NRC Safety Evaluation Report dated April 17, 1986, "Acceptance fo r Referencing of Licensing Topical Report WCAP-8745(P)/8746(NP) , 'Design Bases for the Thermal Overpower 6. T and Thermal Overtemperature 6.T Trip Functions."'

Page 16 of 16

Enclosure II to RA 18-0054 ENCLOSURE II WOLF CREEK GENERATING STATION CYCLE 22 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)

Wolf Creek Generating Station weLFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 22 CORE OPERATING LIMITS REPORT Revision 0 October 2016 Prepared by: 10/3/16 Jeff Blair Date Reviewed by: 10/4/2016 Ian Miller Date Dig itally signed by Gregory 5. Ki nn Approved by:

)4'1 ,

/J; '

DN : cn=Gregory 5. Kinn, o=Wolf Creek, ou=5upervisor Reactor Engineerin g/CD/ Fuel,

- ~ email=grkinn@wcnoc.com, c=U5 Date: 20 16. 10.19 0 1:53:03 -05'00' Gregory S. Kinn Date

Page 1 of 16 DC12 10 /26/ 2016

Wolf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 22 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 (F0 (z)) (Fa Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F! )

3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control 3.3.1 3.4.1 (RAOC) Methodology)

Reactor Trip System (RTS) Instrumentation 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

Wolf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 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 Unacceptabl e Ope ration 660 I

2400 p s ia

--- _I 640

~

'Z--

C>

I-Q)

UJ 2250 p s ia UJ 620 Q)

>Q)

c ca

3::

..2 cl:

600 A cceptabl e Ope ration 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

Wolf Creek Generating Station W$LFCREEK 'NUCLEAR OPERATING CORPORATION Cycle 22 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 I I UNACCEPT'}BLE bPERATUi)N ii:'

~

E 6 .0 , 70%

g_

6 z

C,)

u:

u.

w 0

C,)

w ai::

...cc

, 4 ai:: A r_CEPTAB1j-E w

D,. OPERATION

E w

ai::

0 CC 2 - -

ai::

w Cl 0

E 0

0 10 20 30 40 50 60 70 80 90 100

% of RATBlTHERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs.

THERMAL POWER (%)

Page 4 of 16

Wolf Creek Generating Station W~ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 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) 220 200

. 70~

t) 180 I

( 100% . 1 1 )

s T

E 140 p

s W 120 I

T H 100 D

R A 80 w

N 60 40 20

( 30 2o/o 0) 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~ 222 and :5: 231 steps withdrawn .

Page 5 of 16

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 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 O/o 0

F 90 R

A T

E D80 T

H ACCEPTABLE OPERATION E

R70 M

A L

P50 0

w E

R 50

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

-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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (Fo(ZJ){Fa Methodology) (LCO 3.2 .1, SR 3.2.1.2)

FQ(Z) ~ CFQ *K(Z), f or P > 0.5 p

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

RATED THERMAL POWER CFQ = pRTP Q

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

Q

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

FQM (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 .

Measurement uncertainty is applied as follows.

FJ(Z)=FQM(Z)(1.03)(1.05) =Ft (Z)(1.08 I5) when FQM(Z) is obtained from MIDS.

FJ (Z) = FQM (Z)(l.03 )(UQu) when FQM (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 Uau factor, and it is determined by PDMS .

F; (Z)=FJ (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 FJ (Z) that is determined from an FQM (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 W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 1.2

~

N 1.0 et::

0 I-u<( 0.8 LL

(!)

z

~

<( 0.6 w

CL C

w

~ 0.4

...J

<(

E et::

0 z 0.2 Elevation (ft) K(Z) 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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor ( F; ) (LCO 3.2.2)

Fi shall be limited by the following relationship:

Fi ~Fir [1.0 +PFMl(l.O -P) ]

Where, F:;p = Fi limit at RATED THERMAL POWER (RTP)

= 1.650 PFMl = power factor multiplier for Fi

= 0.3 p = THERMAL POWER RATED THERMAL POWER

= Fi is the measured value of Fi, 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 Fi is obtained from MIDS, the measured value is multiplied by 1.04.

When Fi 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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 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 Value Overtemperature ~T reactor trip setpoint K1 = 1.10 Overtemperature ~T reactor trip setpoint T avg K2 = 0.0137/°F coefficient Overtemperature ~ T reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal T avg at RTP T' ~ 586.5°F Nominal RCS operating pressure P' 2 2235 psig Measured RCS ~T lead/lag constant 11 = 6 sec 12 = 3 sec Measured RCS ~T lag constant 13 = 2 sec Measured RCS average temperature lead/lag 14 = 16 sec constant 15 = 4 sec Measured RCS average temperature lead/lag 15 = O sec constant 0% of RTP when -23% RTP ~ (q.-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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 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 Value Overpower ~ T reactor trip setpoint K4 =1.10 Overpower ~ T reactor trip setpoint T avg Ks= 0.02/°F for increasing Tavg rate/lag coefficient = 0/°F for decreasing T avg Overpower ~ T reactor trip setpoint Tavg heatup Ke= 0.00128/°F for T > T" coefficient = 0/°F for T .:=; T" Indicated Tavg at RTP (calibration temperature T" .:=; 586.5°F for ~ T instrumentation)

Measured RCS ~ T lead/lag constant 1"1 = 6 sec 1"2 = 3 sec Measured RCS ~ T lag constant 1"3 = 2 sec Measured RCS average temperature lead/lag 1"e = O sec constant

Measured RCS average temperature rate/lag constant h (~I) = 0% RTP for all ~I 1"1 = 1O sec

Page 11 of 16

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 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 T avg :S 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% ~k/k) .

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

Safety Analysis DNBR Limit WRB-2 Design Limit DNBR

1. 76 1.23

Page 12 of 16

Wolf Creek Generating Station W$LFCREEK 'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 APPENDIX A A. Input relating to LCO 3.2.1:

F ( Z) max transient l W ( Z) = FQQ

( z teady state x-p, for P > 0.5 F ( Z) max transient l W(Z) = Q x- , for P ~ 0.5 FQ(zteactystate 0 _5 THERMAL POWER where, P =

RATED THERMAL POWER F Q(Zl'at transient = Maximum (F Q(Z) X p) calculated over the entire range of power shapes analyzed for Condition I operations (p = power at which maximum occurs).

FQ(zfeadystate = (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 F 0 (zyteady 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$LFCREEK

' NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0 Input relating to SR 3.2.1.2 Cycle Burnup FQ(Z) Penalty Factor (MWD/MTU) (%)

~ O to :5 7861 2.00 8059 2.01 8257 2.24 8454 2.43 8652 2.53 8850 2.36 9047 2.15

~ 9245 2.00 FQ(Z) Exclusion Zone

(% [INCORE mesh points])

Cycle Burnup (MWD/MTU) Top Bottom

5 8,000 15 [11] 15 [11]

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

Page 14 of 16

Wolf Creek Generating Station weLFCREEK'NUCLEAR OPERATING CORPORATION Cycle 22 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 NRG, 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-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."

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

4.

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.

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).

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

Wolf Creek Generating Station W!LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report Revision 0

7. 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)."

8. 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. "

9. 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). "

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

September 1986.

N RC 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-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, Append ix B, Addendum 1, 'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO.

M86416). "

12. 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)."

13. 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 ~T and Thermal Overtemperature ~T Trip Functions."'

Page 16 of 16

NUCLEAR OPERATING CORPORATION Cynthia R. Hafenstine Manager Nuclear and Regulatory Affairs April 29, 2018 RA 18-0054 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington , DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 22 and Cycle 23 Core Operating Limits Report To Whom It May Concern :

Enclosure I is Revision O of the Wolf Creek Generating Station (WCGS) Cycle 23 Core Operating Limits Report (COLR) . This document is being submitted pursuant to Section 5.6.5 of the WCGS Technical Specifications.

Enclosure II is Revision O of WCGS Cycle 22 COLR. During preparation of Cycle 23 WCGS COLR, it was identified that the Cycle 22 WCGS COLR was not submitted pursuant to Section 5.6.5 of the WCGS Techn ical Specifications. This has been captured in the Corrective Action Program .

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

Sincerely, trn}ftlL re JJ{fw11 Cynth ia R. Hafenstine

)v_,

CRH/rlt Enclosure I - WCGS Cycle 23 Core Operating Limits Report Enclosure II - WCGS Cycle 22 Core Operating Limits Report cc: K. M. Kennedy (NRC), w/e B. K. Singal (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/HC/VET

Enclosure I to RA 18-0054 ENCLOSURE I WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT Revision 0 April 2018 Prepared by: ~ ~ 4/24/2018 Ian Miller Date Reviewed by: /7~

Dustin Wi rth

~ 04/24/2018 Date Approved by: ~1~ 04/24/2018 Gregory S. Kinn Date Page 1 of 16

Wolf Creek Generating Station W$LFCREEK 1

NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 23 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 (F0 (z)) (F a 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 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

Wolf Creek Generating Station W$ LFCREEK 'NUCLEAR OPERATING CORPORATION Cycle 23 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 Unaccept a bl e Operation 660 - - --

- -' - 2400 p s ia

~------I'-- '

r----...__

- ' -~

/

6 40 -........_ -' '

2 000 p s ia U:::-

':!..- / ------- ~ ;--...,.

........__ r---_

~

Cl

~

I---""'" - '*

i-:'

.,,.,, r--.. 225 0 p s ia

~ ' *.

Q)

Q) 620 I'--..

~ ~ ------ ~

\

Q) 1925 p s ia

I'-- \.

' ' \. .

0 cu

3: r-----..

..2

<t: ----r----.. ---- ~

~

\

\.

600

--r--._ \. '

Acceptable Operation '\\

~

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

Wolf Creek Generating Station W lLFCREEK 'NUCLEAR OPERATING CORPORATION Cycle 23 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 pcml°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 UII ACCEPTJ BLE DPERATI< ~N E 6 .0 , 70%

u C.

- 6 z

u ii:

II.

w 0

u w

a:

, 4 1-4 a: A<tCEPTAB E w

a. CPERATIO~

w l-a:

0 1-4 2 a:

w Q

0

ii:

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

% of RA TED T HERM AL POWER Figure 2.2 Moderator Temperature Coefficient Vs.

THERMAL POWER(%)

Page 4 of 16

Wolf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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) 2 20 *( 2 1 . 7 uA , 2 tL ) *( 7 1 .7 0/c . 2, 2) i,

/ ~

200 / /

V V

,/

~A ....... /

180

,/v a ~v V V 160

,t7

( o ob , 1 6 1 )

f-V ( 1 p o °/c

1E 1 )

s T

,- I/ I---- V E 140 p

/e AN Ii.< /

/ C s

W 120 V

/v V

>~/

T H 100

,V

/ t--

~

V t----

D ,__ V ,-

V V

~ 1-- ,-

R A

w 80 l/E AN K L

[

N - / *- - - - - - ,_ ,-. ,_ - ,-

60 / ,/

/ /

V 1(0 ~ >. 4 3 )

40 /

V 20 V - ,- -

- - V - - ~- - - -

0 30 2 °/o 0) [7 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 w ithdrawn shall be the condition where control banks are at a position within the interval of ~ 222 and ~ 231 steps withdrawn.

Page 5 of 16

Wolf Creek Generating Station Wel.FCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 2.5 AXIAL FLUX DIFFERENCE (AFD) (Re laxed 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 O/o 0

F 90 R

A T

E D80 T

ACCEPTABLE H

OPERATION E

R70 M

A L

p 60 0

w E

R 50

( -2 9 , 50) ( 24 , 50) 40

-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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (F0 (Z))(Fo Methodology) (LCO 3.2.1, SR 3.2.1.2)

F0 (Z) :s; CFQ *K(Z), for P > 0.5

- p FQ(Z) :s; c;; *K(Z), for P :s; 0.5 THERMAL POWER where , P =

RA TED THERMAL POWER CFQ = FQ RTP FQ R7P

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

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

FQM (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 .

Measurement uncertainty is applied as follows .

Ff (Z)=F;1 (Z)(l.03)( 1.05)=F;1(Z)(1.0815) when FQM (Z) is obtained from MIDS.

Fi *(Z) = F;1 (Z)(l.03)(UQu ) when F;1 (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 Uau factor, and it is determined by PDMS.

F~v(Z)=Fi °(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/f (Z) uses Ff (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 Wft.FCREEK'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 1.2 N 1.0 S2' 0:::

0

(.)

<(

0.8 ~ - - ------- - - - - - - - '- - - - - - -

LL c.,

z

si::: '

<( 0.6 ,_ ----------

w a.

Q w

N 0.4 ~ -- -

i

<(

~

0:::

0 z 0 .2 ~ Bevation (ft) K(Z) 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

Wolf Creek Generating Station W$ LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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 :

N FM{~ FM,R71' [1.0 + PFM{ ( 1.0- P) ]

Where , Fdi1' = F; limit at RATED THERMAL POWER (RTP)

= 1.650 PFM{ = power factor multiplier for F;

= 0.3 p = 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

Wolf Creek Generating Station W$LFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 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 Value Overtemperature ~ T reactor trip setpo int K1=1.1 0 Overtemperature ~ T reactor trip setpoint T avg K 2 = 0.0137/°F coefficient Overtemperature ~T reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal T avg at RTP T' ~ 586 .5°F Nominal RCS operating pressure P' ~ 2235 psig Measured RCS ~T lead/lag constant -c1 = 6 sec

-c2 = 3 sec Measured RCS ~T lag constant -c3 = 2 sec Measured RCS average temperature lead/lag *4 = 16 sec constant -cs= 4 sec Measured RCS average temperature lead/lag *6 = 0 sec constant 0% of RTP when -23% RTP ~ (q1-qb) ~ 5% RTP Where , q 1 and qb are percent RTP in the upper and lower halves of the core ,

respectively, and q 1 + qb is the total THERMAL POWER in percent RTP.

Page 10 of 16

Wolf Creek Generating Station

\NeLFCREEK

'NUCLEAR OPERATING CORPORATION Cycle 23 Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower Ll T Setpoint Parameter Values (LCO 3.3.1 ,

Table 3.3.1-1 , Note 2)

Parameter Value Overpower Ll T reactor trip setpoint K4 = 1.10 Overpower Ll T reactor trip setpoint T avg K 5 = 0.02/°F for increasing T avg rate/lag coefficient = 0/°F for decreasing T avg Overpower T reactor trip setpoint T avg heatup K 6 = 0.00128/°F for T > T" coefficient = 0/°F for T ~ T" Indicated T avg at RTP (calibration temperature T" ~ 586 .5°F for Ll T instrumentation)

Measu red RCS LlT lead/lag constant *1 = 6 sec

2 = 3 sec Measured RCS LlT lag constant *3 = 2 sec Measured RCS average temperature lead/lag *6 = 0 sec constant Measured RCS average temperature rate/lag *7 = 10 sec constant f 2{Lll) = 0% RTP for all Lll Page 11 of 16