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{{#Wiki_filter:NUCLEAR OPERATING CORPORATION Cynthia R. Hafenstin e Manager Nuclear and R e g u lat o ry Affairs U.S. Nuclear Regulatory Commiss i on ATTN: Document Control Desk Washington , DC 20555 April 29 , 2018 RA 18-0054
{{#Wiki_filter: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:==
==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 i cal 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 , tr n}ftlL re JJ{fw11 )v_, Cynth i a R. Hafenstine 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 Sen i or Res i dent Inspector (NRC), w/e P.O. Box 411 / Burlington , KS 66839 / Phone: (620) 364-8831 An Equal Opportun i ty 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)
Docket No. 50-482: Wolf Creek Generating Station Cycle 22 and Cycle 23 Core Operating Limits Report To Whom It May Concern :
W$LFCRE E K Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT Revision 0 April 2018 Prepared by: 4/24/2018 I an Miller Date Reviewed by: /7~ 04/24/2018 Dustin Wi rt h D ate Approved by: 04/24/2018 Gregory S. Kinn D ate Page 1 of 16 W$LFCREEK 1 NUCLEAR OPERATING CORPORATION 1.0 CORE OPERATING LIMITS REPORT Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 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
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.
: 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  
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 .
-MODE 2 3.2.1 Heat Flux Hot Channel Factor (F 0 (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 W$LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERAT I NG CORPORATION 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 Una ccep t a bl e Operation 660 -------' --' -24 00 p s ia --' / ~-------' -~ I'----' -r----...__  
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
-' ' -------' ' ' 2 000 p s i a ------/ ;--...,. . 6 40 U:::-':!..-........__ I---""'" -r---_ '* ------225 0 p s i a ------' I'--.. r--.. *. --' -----\ . ----------I'--\. ', 1 925 p s i a -----' r-----.. ' ' ' . Cl > i-:' Q) .,, .,, 620 Q) > Q) ::0 cu :3: -------' ' r----.. \. ' ------\ -\. ' --r--._ \. '\\ Acceptable Operation ..2 <t: 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16 Wl LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERAT I NG CORPORATION Core Operating Limits Report Revision 0 2.2 Moderator Temperature Coeff i cient (MTC) (LCO 3.1.3, SR 3.1.3.2) The MTC shall be less positive than the l imit provided in Figure 2.2. The MTC shall be less negative than -50 pcm/&deg;F. The 300 PPM MTC Surveillance limit is -41 pcm l&deg;F (equilibrium, all rods withdrawn , RATED THERMAL POWER condition). The 60 PPM MTC Surveillance limit is -46 pcm/&deg;F (equilibrium , all rods withdrawn, RATED THERMAL POWER condition). '!..... E u C. 8 ;:-6 z !!:! u ii: II. w 0 u w a: ::, 4 1-4 a: w a. ::!:: w l-a: 0 1-4 2 a: w Q 0 :ii: 0 0 10 20 UII ACCEPTJ BLE DPERATI< ~N 6.0 , 70% A<tCEPTAB E CPERATIO~
                                                                                                      )v_,
30 40 50 60 70 80 90 % of RA TED T HERM AL POWER Figure 2.2 Moderator Temperature Coefficient Vs. THERMAL POWER(%) Page 4 of 16 100 W$LFCREEK 'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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).
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
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, 200 180 160 s T E 140 / / ,/ V V ~A ....... / ,/v a ~v V V , t7 f-V ( 1 p o&deg;/c
 
* 1E ( o o b , 1 6 1 ) I/ ,-I----/e AN I i.< / p s W 120 T / C / /v >~ V V / t--V t----V H 100 D R A 80 w N 60 , V V ,__ V ,-1--l/E ,-AN K / L [ -*------,_ ,-. ,_ -,-/ / , / / 40 1(0~ / V >. 4 3) V ----*-,-V -,--20 0 --V ---[7 --~-30 2&deg;/o 0) -0 2 0 40 60 8 0 (FULLY INSERTED)
Enclosure I to RA 18-0054 ENCLOSURE I WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)
THERMAL POWER (Percent)
 
Figure 2.4 Control Bank Insertion, Sequence , and Overlap Limits Vs. THERMAL POWER(%) -Four Loop Operation Full y w it hdrawn shall be the condition where control banks are at a position within the interval of~ 222 and~ 231 steps withdrawn. Page 5 of 16 1 ) V 100 W el.FCREEK W ol f Creek Generating Station Cycle 23 C ore Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 2.5 AXIAL FLUX DIFFERENCE (AFD) (Re l axed Ax i al Offset Control (RAOC) Methodo l ogy) (LCO 3.2.3) The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space is defined by Figure 2.5. 110 100 UNACCEPTABLE O/o OPERATION 0 F 90 R A T E D80 T H E R70 M A L p 60 0 w E R 50 ( -2 9 , 50) 40 30 ( -15 , 100 ) (5 , 100) UNACCEPTABLE OPERATION ACCEPTABLE OPERATION ( 24 , 50) 10 0 10 20 AXIAL FLUX DIFFERENCE
Wolf Creek Generating Station W$LFCREEK
(%AI) Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER(%) Page 6 of 16 30 40 W$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~
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (F 0 (Z))(F o Methodology) (LCO 3.2.1, SR 3.2.1.2) F 0 (Z) :s; CFQ *K(Z), for P > 0.5 -p F Q (Z) :s; c;; *K(Z), for P :s; 0.5 where , P THERMAL POWER = RA TED THERMAL POWER CFQ = RTP F Q R7P F Q = F Q (Z) limit at RATED THERMAL POWER (RTP) = 2.50 , and K (Z) = as defined in Figure 2.6. F QM (Z) is the measured value of F Q (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. F f (Z)=F;1 (Z)(l.03)(1.05)=F;1(Z)(1.0 8 15) when F QM (Z) is obtained from MIDS. F i*(Z) = F;1 (Z)(l.03)(U Qu) when F;1 (Z) is obtained from PDMS. Manufacturing tolerances are accounted for in the 1.03 Engineering uncertainty factor. Measurement uncertainty for M I DS is accounted for in the 1.05 factor. PDMS measurement uncertainty is accounted for in the U a u factor , and it is determined by PDMS. F~v (Z)=F i&deg;(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 F f (Z) that is determined from an F{ (Z) that reflects full-power steady-state conditions rather than current conditions.
Dustin Wi rth
See Appendix A for: F Q Penalty Factor. Page 7 of 16 Wft.FCREEK  
                                        ~                04/24/2018 Date Approved by: ~1~                                04/24/2018 Gregory S. Kinn                   Date Page 1 of 16
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 1.2 N 1.0 S2' 0::: 0 .... (.) <( LL c., z :::si::: <( w a. Q w N ::::i <( 0::: 0 z 0.8 --0.6 ,_ 0.4 ---0.2 0.0 0 ' ---------------------' ' ' ' ----------
 
----' ' Bevation (ft) K(Z) 0.0 1.000 6.0 1.000 12.0 0.925 2 4 6 8 10 12 CORE HEIGHT (FT) Figure 2.6 K(Z) -Normalized Peaking Factor Vs. Core Height Page 8 of 16 W$LFCREEK 'NUCLEAR OPERAT I NG CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor ( F;) (LCO 3.2.2) F; shall be lim i ted by the following relationship
Wolf Creek Generating Station W$LFCREEK 1
: N R71'[ ( )] F M{~ F M , 1.0 + PF M{ 1.0-P Where , Fdi 1' = F; limit at RATED THERMAL POWER (RTP) = 1.650 PF M{ = 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 appl i ed 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 dete r mined by PDMS , with a lower limit of 4%. Page 9 of 16 W$LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 2.8 Reactor Tr i p System Overtemperature T Setpoint Parameter Values (LCO 3.3.1 , Table 3.3.1-1 , Note 1) Parameter Overtemperature T reactor trip setpo i nt Overtemperature T reactor trip setpoint T avg coefficient O v ertemperature T reactor trip setpoint pressure coefficient Nominal T avg 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 average temperature lead/lag constant Value K 1=1.1 0 K 2 = 0.0137/&deg;F K 3 = 0.000671/psig T' 586.5&deg;F P' 2235 psig -c1 = 6 sec -c2 = 3 sec -c3 = 2 sec *4 = 16 sec -cs= 4 sec *6 = 0 sec 0% of RTP when -23% RTP (q 1-q b) 5% RTP Where , q 1 and q b are percent RTP in the upper and lower halves of the core , respectively, and q 1 + q b is the total THERMAL POWER in percent RTP. Page 10 o f 16
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 .
\Ne LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERATING CORPORATION 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 Overpower Ll T reactor trip setpoint Overpower Ll T reactor trip setpoint T a v g rate/lag coefficient Overpower T reactor trip setpoint T a v g heatup coefficient Indicated T a v g at RTP (calibration temperature for Ll T instrumentation)
The core operating limits that are included in the COLR affect the following Technical Specifications :
Measu r ed RCS Ll T lead/lag constant Measured RCS Ll T lag constant Measured RCS average temperature lead/lag constant Measured RCS average temperature rate/lag constant f 2{Ll l) = 0% RTP for all Ll l Page 1 1 of 16 Value K 4 = 1.10 K 5= 0.02/&deg;F for increasing T a v g = 0/&deg;F for decreasing T a v g K 6 = 0.00128/&deg;F for T > T" = 0/&deg;F for T T" T" 586.5&deg;F *1 = 6 sec *2 = 3 sec *3 = 2 sec *6 = 0 sec *7 = 10 sec W$LFCREEK
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! )
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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 &deg;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% Li k/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 W$LFCREEK Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION APPENDIX A A. Input relating to LCO 3.2.1: Fa ( Z) max tran s ie nt } W(Z)= -x-forP> 0.5 F g (Z)51 ead ys tate p , F ( Z) max tran s i e nt I W(Z)-0 d x 0_5 , forP :S 0.5 F Q (Zft ea ys tate where , P= T HERMAL POWER RATED T H ERMAL POW E R F o(Z)""'' = Maximum (F Q(Z) x p) calculated over the entire range of power shapes -lr a nsi e nr analyzed for Condition I operations (p = power at which maximum occurs). F g (Zf ea d y s l a l e = (F Q(Z) x p) calculated at full power (p = 1.0) equilibrium conditions. The W(z) values are generated at full power equilibrium conditions (P = 1.0). W(z) values specific to part-power conditions may also be generated; these can be used for part-power surveillance measurements , rathe r than the full-power W(z) values. For these part-power W(z) va l ues , the F Q (z f t e ady s t ate (denominator in above equations) is generated at the specific anticipated surveillance conditions. W(Z) values are issued in controlled reports which will be provided on request. Page 13 o f 16 W$LFCREEK Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION Input relating to SR 3.2.1.2 Cycle Burnup (MWD/MTU) 0 to :5 7658 7856 8053 8251 8449 8646 8844 9041 9239 9437 Cycle Burnup (MWD/MTU)
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:
:5 8 , 000 > 8 , 000 F Q (Z) Penalty Factor (%) 2.00 2.14 2.37 2.63 2.86 2.79 2.57 2.32 2.06 2.00 F Q (Z) Exclusion Zone (% [INCORE mesh points]) Top Bottom 15 [11] 15 [11] 10 [7] 10 [7] Page 14 of 16 W$LFCREEK 'NUCLEAR OPERATING CORPORATION Wo l f Creek Generating Station 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 Hyd r aulic Analysis Methodology for the Wolf Creek Generating Station." 2. WCAP-11397-P
ASA     B 3.4.1 RCS Pressure, Temperature , and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 16
-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 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 1 A , " 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 , Re l axation of Constant Axial Offset Control -F a Surveillance Technical Specification" (TAC No. M88206). 5. WCNOC Topical Report NSAG-007 , " Re l oad 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 W$LFCREEK  
Wolf Creek Generating Station W$ 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:
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 7. WCAP-16009-P-A , " Real i s t ic Large B r eak LOCA Evaluat i on Methodology Using Automated Statistical Treatment of Uncertain t y Method (ASTRUM)," Revision 0 , Janua r y 2005. NRC letter dated November 5 , 2004 ," Final Safety Evaluation for WCAP-16009
2.1   Reactor Core Safety Limits (SL 2.1 .1)
-P , Rev i sion 0 , " Realistic Large Break LOCA Evaluation Methodology Using Au t omated Stat i stical Treatment of Uncertainty Method (ASTRUM)" (TAC NO. MB9483)." 8. WCAP-16045-P-A , " Qualification of the Two-D i mensional Transport Code PARAGON ," August 2004. NRC Safety Evaluation dated Ma r ch 18 , 2004 , " Final Safety Evaluation for Wes ti nghouse Topical Report WCAP-16045-P , Revisio n 0 , " Qualification of the Dimensional Transport Code PARAGON." 9. WCAP-16045-P-A , Addendum 1-A , " Qualification of the NEXUS Nuc l ear Data Methodology
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.
," August 200 7. NRC Safety Evaluation dated February 23 , 2007 , " Final Safety Evaluation fo r Westinghouse Electric Company (Westinghouse)
680 Unaccept a bl e Operation 660             - - --
Topical Report (TR) WCAP-16045-P-A , Addendum 1 , " Qual i fication of the NEXUS Nuclear Data Methodology" (T AC NO. MC9606)." 10. WCAP 10965-P-A , " ANC: A West i nghouse Advanced Nodal Computer Code ," September 1986. NRC letter dated June 23 , 1986 , " Accep t ance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP." 11. WCAP-12610-P-A , " VANTAGE+ Fuel Assembly Reference Core Report ," Ap ri 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
                                                    - -' -                                   2400 p s ia
-126 1 0 , Appendix B , Addendum 1 , 'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M864 1 6)." 12. WCAP-12610
                        --
-P-A & CENPD-404-P-A , Addendum 1-A , " Optimized Zirlo&#x17d;," July 2006. NRC Safety Evaluation dated June 10 , 2005 , " F inal Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A , " Optim i zed Zirlo&#x17d;," (TAC NO. MB8041)." 1 3. WCAP-87 45-P-A , " Design Bases for the Thermal Overpower
                                ~------I'--                   '
: 6. T and Thermal Overtemperature
r----...__
: 6. T Trip Function." September 1986. NRC Safety Evaluation Report dated April 1 7 , 1 9 8 6 , " Acceptance fo r Referencing of Licens i ng Topical Report WCAP-8745 (P)/8746 (NP), 'Design Bases for the The r mal Overpower
                                                                      - ' -~
: 6. T and Therma l Overtemp e rature 6. T Trip Functions."' Page 16 o f 16 Enclosure II to RA 18-0054 ENCLOSURE II WOLF CREEK GENERATING STATION CYCLE 22 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)
                                                                                    --
* *
                                                                                        /
* weLFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION CYCLE 22 CORE OPERATING LIMITS REPORT Prepared by: Reviewed by: Approved by: Revision 0 October 2016 10/3/16 Jeff Blair Date 10/4/2016 Ian Miller Date )4'1 Dig ita ll y signed by Gregory 5. Ki nn , DN: cn=Gregory
                                                                                          ' -' - -
: 5. Kinn, o=Wo lf C reek , /J; ' ou=5 up erviso r Reactor En g in ee rin g/C D/Fuel, * -~ e mail=g r kinn@wcnoc.com, c=U5 Dat e: 20 1 6.10.1 9 0 1:53: 0 3 -05'00' Gregory S. Kinn Date DC12 10/26/2016 Page 1 of 16 
                                                                                                      ' -
* *
6 40                                                  -........_                               -' '
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 22 has been prepared in accordance with the requirements of Technical Specification 5.6.5. The core operating limits that are included in the COLR affect the following Technical Specifications
                                                                                                                            -- .
: 2.1.1 Reactor Core Safety Limits 3.1.1 Shutdown Margin (SOM) 3.1.3 Moderator Temperature Coefficient (MTC) 3. 1 .4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 PHYSICS TESTS Exceptions  
                        ----
-MODE 2 3.2.1 3.2.2 3.2.3 3.3.1 3.4.1 Heat Flux Hot Channel Factor (F 0 (z)) (F a Methodology)
                                                                                                                    '
Nuclear Entha l py Rise Hot Channel Factor (F!) AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology)
2 000 p s ia
Reactor Tr i p System (RTS) Instrumentation RCS Pressure, Temperature , and Flow Departure from Nucleate Boiling (DNB) Limits 3.9.1 Boron Concentrat i on The portions of the Technical Specif i cation Bases affected by the report are listed below: ASA B 3.4.1 RCS Pressure , Temperature , and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 16
                                                                                                                        ' '
* *
U:::-
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 '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. 680 660 640 'Z--C> > .,, I-Q) UJ UJ 620 Q) > Q) :c ca 3:: ..2 cl: 600 580 560 0.0 I 24 00 p s i a ---_ I 225 0 p s i a A cce ptabl e Op e ration 0.2 0.4 0.6 0.8 Un acce ptabl e Op e ration " ' ' . . 1.0 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16 ' ' 1.2 
    ':!..-                                      /                              ------- ~ ;--...,.
* *
                                                                  -- -----
* W$LFCREEK Wolf Creek Generating Station Cycle 22 'NUCLEAR OPERATING CORPORATION 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/&deg;F. The 300 PPM MTC Surveillance limit is -41 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
                        ........__                                                                           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/&deg;F.
The 300 PPM MTC Surveillance limit is -41 pcml&deg;F (equilibrium, all rods withdrawn , RATED THERMAL POWER condition) .
The 60 PPM MTC Surveillance limit is -46 pcm/&deg;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 &deg;/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 &deg;/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 &deg;(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/&deg;F coefficient Overtemperature     ~T reactor trip setpoint pressure      K3 = 0.000671/psig coefficient Nominal T avg at RTP                                        T'  ~ 586 .5&deg;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/&deg;F for increasing T avg rate/lag coefficient                                   = 0/&deg;F for decreasing T avg Overpower T reactor trip setpoint T avg heatup     K 6 = 0.00128/&deg;F for T > T" coefficient                                           = 0/&deg;F for T ~ T" Indicated T avg at RTP (calibration temperature   T"  ~ 586 .5&deg;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 &deg;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$LFCREEK  'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report
* Revision 0 2.2       Moderator Temperature Coefficient (MTC) (LCO 3.1 .3, SR 3.1.3.2)
The MTC shall be less positive than the limit provided in Figure 2.2.
The MTC shall be less negative than -50 pcm/&deg;F.
The 300 PPM MTC Surveillance limit is -41 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
ii:' E g_ 8 ;: 6 z !!! C,) u: u. w 0 C,) w ai:: :, 4 ... cc ai:: w D,. :E w ... ai:: 0 ... CC 2 ai:: w Cl 0 :E 0 --0 1 0 20 I I UNACCEPT'}BLE bPERATUi)N 6.0 , 70% A r_CEPTAB1j-E OPERATION 30 40 50 60 70 80 90 % of RATBlTHERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs. THERMAL POWER (%) Page 4 of 16 100 
8 I      I UNACCEPT'}BLE bPERATUi)N ii:'
* *
      ~
* W~LFCREEK 'NUCLEAR OPERAT I NG CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 2.3 Shutdown Bank Insertion Limits (LCO 3.1.5) The shutdown banks shall be fully withdrawn (i.e., positioned within the interval of 222 and 231 steps withdrawn).
E                                                         6 .0 , 70%
2.4 Control Bank Insertion Limits (LCO 3.1.6) The Control Bank insertion , sequence, and overlap limits are specified in Figure 2.4. (FULLY WITHDRAWN) 220 200 .70~
g_
* t) I 180 ( 100% . 1 1 ) s T E 140 p s W 120 I T H 100 D R A 80 w N 60 40 20 0 ( 30 2o/o 0) 0 20 40 60 80 (FULLY INSERTED)
      ;: 6 z
THERMAL POWER (Percent)
      !!!
Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs. THERMAL POWER (%) -Four Loop Operation Fully withdrawn shall be the condition where control banks are at a position within the interval of~ 222 and :5: 231 steps withdrawn. Page 5 of 16 100 
C,)
* *
u:
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 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 O/o 0 F 90 R A T E D80 T H E R70 M A L P50 0 w E R 50 40 -40 UNACCEPTABLE OPERATION ( -29 , 50) -30 ( -15 , 100 ) (5 , 100) UNACCEPTABLE OPERATION ACCEPTABLE OPERATION ( 24 , 50) 10 0 10 20 AXIAL FLUX DIFFERENCE
u.
(%A I) Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER (%) Page 6 of 16 30 40 
w
* *
* 0 C,)
* W$LFCREEK  
w ai::
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (Fo(Z J){F a Methodology) (LCO 3.2.1, SR 3.2.1.2) FQ(Z) CF Q *K(Z), f o r P > 0.5 p FQ(Z) c;; *K(Z), for P 0.5 where , P = CFQ = pR TP = Q THERMAL P O WE R RATE D THERMAL PO WE R pRTP Q F Q(Z) limit at RATED THERMAL POWER (RTP) = 2.50, and K(Z) = as defined in Figure 2.6. F QM (Z) i s the measured value of F Q(Z), inferred from a power distribution measurement obtained with the Movable lncore Detector System (MIDS) or the Power Distribution Monitoring System. Measurement uncertainty is applied as follows. F J(Z)=F QM(Z)(1.03)(1.
      ...cc
0 5)=F t(Z)(1.08 I 5) when F QM(Z) i s obta i ned from MIDS. F J (Z) = F QM (Z)(l.03 )(U Qu) when F QM (Z) is obtained from PDMS. Manufacturing tolerances are accounted for in the 1.03 Engineering uncertainty factor. Measurement uncertainty for MIDS is accounted for in the 1.05 factor. PDMS measurement uncertainty is accounted for in the Uau factor , and it is determined by PDMS. F; (Z)=FJ (Z) W(Z) where, W(Z) = a cycle dependent function that accounts for power distribution transients encountered during normal operation (see Appendix A). When using the PDMS , F; (Z) uses F J (Z) that is determined from an F QM (Z) that reflects full-power steady-state conditions rather than current conditions . See Appendix A for: FQ Penalty Factor. Page 7 of 16
:,    4 ai::                       A r_CEPTAB1j-E w
* *
D,.                         OPERATION
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 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
:E
* *
      ...
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 2.7 Nuclear Enthalpy Rise Hot Channel Factor ( F;) (LCO 3.2.2) Fi shall be limited by the following relationship:
w ai::
Fi ~Fir[1.0+PFMl(l.O-P)] Where , F:;p = Fi limit at RA TED THERMAL POWER (RTP) = 1.650 PF Ml = power factor multiplier for F i p = 0.3 = THE R MAL POWE R RATED THE R MAL PO WER = F i is the measured value of Fi, inferred from a power distribution measurement obta i ned with the Movable lncore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS). Measurement uncertainty is applied as follows. When Fi is obtained from MIDS , the measured value is multiplied by 1.04. When Fi is obtained from PDMS, the measured value is increased by an uncertainty factor (U~H), and the factor is determined by PDMS , with a lower limit of 4% . Page 9 of 16
      ...
* *
0 CC     2   -    -
* W$LFCREEK Wolf Creek Generating Station Cycle 22 '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 Overtemperature T reactor trip setpoint Overtemperature T reactor trip setpoint T avg coefficient Overtemperature T reactor trip setpoint pressure coefficient Nominal T avg 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 average temperature lead/lag constant Value K1 = 1.10 K2 = 0.0137/&deg;F K3 = 0.000671/psig T' 586.5&deg;F P' 2 2235 psig 11 = 6 sec 12 = 3 sec 13 = 2 sec 14 = 16 sec 15 = 4 sec 15 = O sec 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
ai::
* *
w Cl 0
* W$LFCREEK Wolf Creek Generating Station Cycle 22 'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower~ T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1 , Note 2) Parameter Overpower~ T reactor trip setpoint Overpower~ T reactor trip setpoint T avg rate/lag coefficient Overpower T reactor trip setpoint T avg heatup coefficient Indicated Tavg at RTP (calibration temperature for~ T instrumentation)
:E 0
Measured RCS T lead/lag constant Measured RCS T lag constant Measured RCS average temperature lead/lag constant Measured RCS average temperature rate/lag constant h (~I) = 0% RTP for all ~I Page 11 of 16 Value K 4=1.10 Ks= 0.02/&deg;F for increasing Tavg = 0/&deg;F for decreasing T avg Ke= 0.00128/&deg;F for T > T" = 0/&deg;F for T .:=; T" T" .:=; 586.5&deg;F 1"1 = 6 sec 1"2 = 3 sec 1"3 = 2 sec 1"e = O sec 1" 1 = 1 O sec  
0     10    20     30        40      50    60      70      80      90  100
* *
                                          % of RATBlTHERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs.
* W$LFCREEK
THERMAL POWER (%)
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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 &deg;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 1. 76 WRB-2 Design Limit DNBR 1.23 Page 12 of 16  
* Page 4 of 16
* * * . '* W$LFCREEK
 
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 A. APPENDIX A Input relating to LCO 3.2.1: F ( Z) m ax tran s i ent l W ( Z) = Q x -for P > 0.5 FQ ( zt ea d y s tate p, F ( Z) m ax tr a n s ient l W(Z) = Q x-, for P 0.5 FQ(zteact ys t a te 0_5 where , THERMAL POWER P= RA TED THERMAL P OWER F Q(Zl'a t tr ans i e nt = 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(zf e ad ys t a t e = (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 st a te (denominator in above equations) is generated at the specific anticipated surveillance conditions.
Wolf Creek Generating Station W~ LFCREEK
W(Z) values are issued in controlled reports which will be provided on request. Page 13 of 16 
          'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report
* *
* Revision 0 2.3   Shutdown Bank Insertion Limits (LCO 3.1.5)
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION Input relating to SR 3.2.1.2 Cycle Burnup F Q (Z) Penalty Factor (MWD/MTU) O to :5 7861 8059 8257 8454 8652 8850 9047 9245 Cycle Burnup (MWD/MTU)
The shutdown banks shall be fully withdrawn (i.e., positioned within the interval of ~ 222 and ~ 231 steps withdrawn).
:5 8,000 > 8,000 (%) 2.00 2.01 2.24 2.43 2.53 2.36 2.15 2.00 F Q (Z) Exclusion Zone (% [INCORE mesh points]) Top 15 [11] 10 [7] Page 14 of 16 Bottom 15 [11] 10 [7] 
2.4   Control Bank Insertion Limits (LCO 3.1.6)
* *
The Control Bank insertion, sequence, and overlap limits are specified in Figure 2.4.
* weLFCREEK  
(FULLY WITHDRAWN) 220 200
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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).
                                                                                              . 70~
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 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 -Fa Surveillance Technical Specification
* t) 180 I
," 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).
( 100% . 1 1 )
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
s T
* *
E 140
* W!LFCREEK  
* p s
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 7. WCAP-16009-P-A , "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncerta i nty Method (ASTRUM)," Revision 0 , January 2005. NRC letter dated November 5 , 2004 ," Final Safety Evaluation for WCAP-16009-P , Revis i on 0 , " Realistic La r ge 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
W   120 I
-P , Revision 0 , "Qualification of the 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)
T H 100 D
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+
R A   80 w
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 i x B, Addendum 1, 'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M86416)." 12. WCAP-12610-P-A  
N 60 40 20
& CENPD-404-P-A , Addendum 1-A , " Optimized Zirlo&#x17d;," 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&#x17d;," (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  
( 30 2o/o 0) 0 0                20                   40               60                 80                 100 (FULLY INSERTED)                       THERMAL POWER (Percent)
~T and Thermal Overtemperature
Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs.
~T Trip Functions."' Page 16 of 16 NUCLEAR OPERATING CORPORATION Cynthia R. Hafenstin e Manager Nuclear and R e g u lat o ry Affairs U.S. Nuclear Regulatory Commiss i on ATTN: Document Control Desk Washington , DC 20555 April 29 , 2018 RA 18-0054
* 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/&deg;F coefficient Overtemperature ~ T reactor trip setpoint pressure         K3 = 0.000671/psig coefficient Nominal T avg at RTP                                       T' ~ 586.5&deg;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/&deg;F for increasing Tavg rate/lag coefficient                                 = 0/&deg;F for decreasing T avg Overpower ~ T reactor trip setpoint Tavg heatup   Ke= 0.00128/&deg;F for T > T" coefficient                                          = 0/&deg;F for T .:=; T" Indicated Tavg at RTP (calibration temperature   T" .:=; 586.5&deg;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 &deg;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:==
==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 i cal 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 , tr n}ftlL re JJ{fw11 )v_, Cynth i a R. Hafenstine 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 Sen i or Res i dent Inspector (NRC), w/e P.O. Box 411 / Burlington , KS 66839 / Phone: (620) 364-8831 An Equal Opportun i ty 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)
Docket No. 50-482: Wolf Creek Generating Station Cycle 22 and Cycle 23 Core Operating Limits Report To Whom It May Concern :
W$LFCRE E K Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT Revision 0 April 2018 Prepared by: 4/24/2018 I an Miller Date Reviewed by: /7~ 04/24/2018 Dustin Wi rt h D ate Approved by: 04/24/2018 Gregory S. Kinn D ate Page 1 of 16 W$LFCREEK 1 NUCLEAR OPERATING CORPORATION 1.0 CORE OPERATING LIMITS REPORT Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 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
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.
: 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  
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 .
-MODE 2 3.2.1 Heat Flux Hot Channel Factor (F 0 (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 W$LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERAT I NG CORPORATION 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 Una ccep t a bl e Operation 660 -------' --' -24 00 p s ia --' / ~-------' -~ I'----' -r----...__  
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
-' ' -------' ' ' 2 000 p s i a ------/ ;--...,. . 6 40 U:::-':!..-........__ I---""'" -r---_ '* ------225 0 p s i a ------' I'--.. r--.. *. --' -----\ . ----------I'--\. ', 1 925 p s i a -----' r-----.. ' ' ' . Cl > i-:' Q) .,, .,, 620 Q) > Q) ::0 cu :3: -------' ' r----.. \. ' ------\ -\. ' --r--._ \. '\\ Acceptable Operation ..2 <t: 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16 Wl LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERAT I NG CORPORATION Core Operating Limits Report Revision 0 2.2 Moderator Temperature Coeff i cient (MTC) (LCO 3.1.3, SR 3.1.3.2) The MTC shall be less positive than the l imit provided in Figure 2.2. The MTC shall be less negative than -50 pcm/&deg;F. The 300 PPM MTC Surveillance limit is -41 pcm l&deg;F (equilibrium, all rods withdrawn , RATED THERMAL POWER condition). The 60 PPM MTC Surveillance limit is -46 pcm/&deg;F (equilibrium , all rods withdrawn, RATED THERMAL POWER condition). '!..... E u C. 8 ;:-6 z !!:! u ii: II. w 0 u w a: ::, 4 1-4 a: w a. ::!:: w l-a: 0 1-4 2 a: w Q 0 :ii: 0 0 10 20 UII ACCEPTJ BLE DPERATI< ~N 6.0 , 70% A<tCEPTAB E CPERATIO~
                                                                                                      )v_,
30 40 50 60 70 80 90 % of RA TED T HERM AL POWER Figure 2.2 Moderator Temperature Coefficient Vs. THERMAL POWER(%) Page 4 of 16 100 W$LFCREEK 'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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).
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
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, 200 180 160 s T E 140 / / ,/ V V ~A ....... / ,/v a ~v V V , t7 f-V ( 1 p o&deg;/c
 
* 1E ( o o b , 1 6 1 ) I/ ,-I----/e AN I i.< / p s W 120 T / C / /v >~ V V / t--V t----V H 100 D R A 80 w N 60 , V V ,__ V ,-1--l/E ,-AN K / L [ -*------,_ ,-. ,_ -,-/ / , / / 40 1(0~ / V >. 4 3) V ----*-,-V -,--20 0 --V ---[7 --~-30 2&deg;/o 0) -0 2 0 40 60 8 0 (FULLY INSERTED)
Enclosure I to RA 18-0054 ENCLOSURE I WOLF CREEK GENERATING STATION CYCLE 23 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)
THERMAL POWER (Percent)
 
Figure 2.4 Control Bank Insertion, Sequence , and Overlap Limits Vs. THERMAL POWER(%) -Four Loop Operation Full y w it hdrawn shall be the condition where control banks are at a position within the interval of~ 222 and~ 231 steps withdrawn. Page 5 of 16 1 ) V 100 W el.FCREEK W ol f Creek Generating Station Cycle 23 C ore Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 2.5 AXIAL FLUX DIFFERENCE (AFD) (Re l axed Ax i al Offset Control (RAOC) Methodo l ogy) (LCO 3.2.3) The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space is defined by Figure 2.5. 110 100 UNACCEPTABLE O/o OPERATION 0 F 90 R A T E D80 T H E R70 M A L p 60 0 w E R 50 ( -2 9 , 50) 40 30 ( -15 , 100 ) (5 , 100) UNACCEPTABLE OPERATION ACCEPTABLE OPERATION ( 24 , 50) 10 0 10 20 AXIAL FLUX DIFFERENCE
Wolf Creek Generating Station W$LFCREEK
(%AI) Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER(%) Page 6 of 16 30 40 W$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~
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (F 0 (Z))(F o Methodology) (LCO 3.2.1, SR 3.2.1.2) F 0 (Z) :s; CFQ *K(Z), for P > 0.5 -p F Q (Z) :s; c;; *K(Z), for P :s; 0.5 where , P THERMAL POWER = RA TED THERMAL POWER CFQ = RTP F Q R7P F Q = F Q (Z) limit at RATED THERMAL POWER (RTP) = 2.50 , and K (Z) = as defined in Figure 2.6. F QM (Z) is the measured value of F Q (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. F f (Z)=F;1 (Z)(l.03)(1.05)=F;1(Z)(1.0 8 15) when F QM (Z) is obtained from MIDS. F i*(Z) = F;1 (Z)(l.03)(U Qu) when F;1 (Z) is obtained from PDMS. Manufacturing tolerances are accounted for in the 1.03 Engineering uncertainty factor. Measurement uncertainty for M I DS is accounted for in the 1.05 factor. PDMS measurement uncertainty is accounted for in the U a u factor , and it is determined by PDMS. F~v (Z)=F i&deg;(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 F f (Z) that is determined from an F{ (Z) that reflects full-power steady-state conditions rather than current conditions.
Dustin Wi rth
See Appendix A for: F Q Penalty Factor. Page 7 of 16 Wft.FCREEK  
                                        ~                04/24/2018 Date Approved by: ~1~                                04/24/2018 Gregory S. Kinn                   Date Page 1 of 16
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 1.2 N 1.0 S2' 0::: 0 .... (.) <( LL c., z :::si::: <( w a. Q w N ::::i <( 0::: 0 z 0.8 --0.6 ,_ 0.4 ---0.2 0.0 0 ' ---------------------' ' ' ' ----------
 
----' ' Bevation (ft) K(Z) 0.0 1.000 6.0 1.000 12.0 0.925 2 4 6 8 10 12 CORE HEIGHT (FT) Figure 2.6 K(Z) -Normalized Peaking Factor Vs. Core Height Page 8 of 16 W$LFCREEK 'NUCLEAR OPERAT I NG CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor ( F;) (LCO 3.2.2) F; shall be lim i ted by the following relationship
Wolf Creek Generating Station W$LFCREEK 1
: N R71'[ ( )] F M{~ F M , 1.0 + PF M{ 1.0-P Where , Fdi 1' = F; limit at RATED THERMAL POWER (RTP) = 1.650 PF M{ = 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 appl i ed 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 dete r mined by PDMS , with a lower limit of 4%. Page 9 of 16 W$LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 2.8 Reactor Tr i p System Overtemperature T Setpoint Parameter Values (LCO 3.3.1 , Table 3.3.1-1 , Note 1) Parameter Overtemperature T reactor trip setpo i nt Overtemperature T reactor trip setpoint T avg coefficient O v ertemperature T reactor trip setpoint pressure coefficient Nominal T avg 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 average temperature lead/lag constant Value K 1=1.1 0 K 2 = 0.0137/&deg;F K 3 = 0.000671/psig T' 586.5&deg;F P' 2235 psig -c1 = 6 sec -c2 = 3 sec -c3 = 2 sec *4 = 16 sec -cs= 4 sec *6 = 0 sec 0% of RTP when -23% RTP (q 1-q b) 5% RTP Where , q 1 and q b are percent RTP in the upper and lower halves of the core , respectively, and q 1 + q b is the total THERMAL POWER in percent RTP. Page 10 o f 16
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 .
\Ne LFCREEK Wolf Creek Generating Station Cycle 23 'NUCLEAR OPERATING CORPORATION 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 Overpower Ll T reactor trip setpoint Overpower Ll T reactor trip setpoint T a v g rate/lag coefficient Overpower T reactor trip setpoint T a v g heatup coefficient Indicated T a v g at RTP (calibration temperature for Ll T instrumentation)
The core operating limits that are included in the COLR affect the following Technical Specifications :
Measu r ed RCS Ll T lead/lag constant Measured RCS Ll T lag constant Measured RCS average temperature lead/lag constant Measured RCS average temperature rate/lag constant f 2{Ll l) = 0% RTP for all Ll l Page 1 1 of 16 Value K 4 = 1.10 K 5= 0.02/&deg;F for increasing T a v g = 0/&deg;F for decreasing T a v g K 6 = 0.00128/&deg;F for T > T" = 0/&deg;F for T T" T" 586.5&deg;F *1 = 6 sec *2 = 3 sec *3 = 2 sec *6 = 0 sec *7 = 10 sec W$LFCREEK
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! )
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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 &deg;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% Li k/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 W$LFCREEK Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION APPENDIX A A. Input relating to LCO 3.2.1: Fa ( Z) max tran s ie nt } W(Z)= -x-forP> 0.5 F g (Z)51 ead ys tate p , F ( Z) max tran s i e nt I W(Z)-0 d x 0_5 , forP :S 0.5 F Q (Zft ea ys tate where , P= T HERMAL POWER RATED T H ERMAL POW E R F o(Z)""'' = Maximum (F Q(Z) x p) calculated over the entire range of power shapes -lr a nsi e nr analyzed for Condition I operations (p = power at which maximum occurs). F g (Zf ea d y s l a l e = (F Q(Z) x p) calculated at full power (p = 1.0) equilibrium conditions. The W(z) values are generated at full power equilibrium conditions (P = 1.0). W(z) values specific to part-power conditions may also be generated; these can be used for part-power surveillance measurements , rathe r than the full-power W(z) values. For these part-power W(z) va l ues , the F Q (z f t e ady s t ate (denominator in above equations) is generated at the specific anticipated surveillance conditions. W(Z) values are issued in controlled reports which will be provided on request. Page 13 o f 16 W$LFCREEK Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION Input relating to SR 3.2.1.2 Cycle Burnup (MWD/MTU) 0 to :5 7658 7856 8053 8251 8449 8646 8844 9041 9239 9437 Cycle Burnup (MWD/MTU)
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:
:5 8 , 000 > 8 , 000 F Q (Z) Penalty Factor (%) 2.00 2.14 2.37 2.63 2.86 2.79 2.57 2.32 2.06 2.00 F Q (Z) Exclusion Zone (% [INCORE mesh points]) Top Bottom 15 [11] 15 [11] 10 [7] 10 [7] Page 14 of 16 W$LFCREEK 'NUCLEAR OPERATING CORPORATION Wo l f Creek Generating Station 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 Hyd r aulic Analysis Methodology for the Wolf Creek Generating Station." 2. WCAP-11397-P
ASA     B 3.4.1 RCS Pressure, Temperature , and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 16
-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 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 1 A , " 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 , Re l axation of Constant Axial Offset Control -F a Surveillance Technical Specification" (TAC No. M88206). 5. WCNOC Topical Report NSAG-007 , " Re l oad 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 W$LFCREEK  
Wolf Creek Generating Station W$ 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:
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 23 Core Operating Limits Report Revision 0 7. WCAP-16009-P-A , " Real i s t ic Large B r eak LOCA Evaluat i on Methodology Using Automated Statistical Treatment of Uncertain t y Method (ASTRUM)," Revision 0 , Janua r y 2005. NRC letter dated November 5 , 2004 ," Final Safety Evaluation for WCAP-16009
2.1   Reactor Core Safety Limits (SL 2.1 .1)
-P , Rev i sion 0 , " Realistic Large Break LOCA Evaluation Methodology Using Au t omated Stat i stical Treatment of Uncertainty Method (ASTRUM)" (TAC NO. MB9483)." 8. WCAP-16045-P-A , " Qualification of the Two-D i mensional Transport Code PARAGON ," August 2004. NRC Safety Evaluation dated Ma r ch 18 , 2004 , " Final Safety Evaluation for Wes ti nghouse Topical Report WCAP-16045-P , Revisio n 0 , " Qualification of the Dimensional Transport Code PARAGON." 9. WCAP-16045-P-A , Addendum 1-A , " Qualification of the NEXUS Nuc l ear Data Methodology
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.
," August 200 7. NRC Safety Evaluation dated February 23 , 2007 , " Final Safety Evaluation fo r Westinghouse Electric Company (Westinghouse)
680 Unaccept a bl e Operation 660             - - --
Topical Report (TR) WCAP-16045-P-A , Addendum 1 , " Qual i fication of the NEXUS Nuclear Data Methodology" (T AC NO. MC9606)." 10. WCAP 10965-P-A , " ANC: A West i nghouse Advanced Nodal Computer Code ," September 1986. NRC letter dated June 23 , 1986 , " Accep t ance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP." 11. WCAP-12610-P-A , " VANTAGE+ Fuel Assembly Reference Core Report ," Ap ri 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
                                                    - -' -                                   2400 p s ia
-126 1 0 , Appendix B , Addendum 1 , 'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M864 1 6)." 12. WCAP-12610
                        --
-P-A & CENPD-404-P-A , Addendum 1-A , " Optimized Zirlo&#x17d;," July 2006. NRC Safety Evaluation dated June 10 , 2005 , " F inal Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A , " Optim i zed Zirlo&#x17d;," (TAC NO. MB8041)." 1 3. WCAP-87 45-P-A , " Design Bases for the Thermal Overpower
                                ~------I'--                   '
: 6. T and Thermal Overtemperature
r----...__
: 6. T Trip Function." September 1986. NRC Safety Evaluation Report dated April 1 7 , 1 9 8 6 , " Acceptance fo r Referencing of Licens i ng Topical Report WCAP-8745 (P)/8746 (NP), 'Design Bases for the The r mal Overpower
                                                                      - ' -~
: 6. T and Therma l Overtemp e rature 6. T Trip Functions."' Page 16 o f 16 Enclosure II to RA 18-0054 ENCLOSURE II WOLF CREEK GENERATING STATION CYCLE 22 CORE OPERATING LIMITS REPORT, Revision 0 (16 pages)
                                                                                    --
* *
                                                                                        /
* weLFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION CYCLE 22 CORE OPERATING LIMITS REPORT Prepared by: Reviewed by: Approved by: Revision 0 October 2016 10/3/16 Jeff Blair Date 10/4/2016 Ian Miller Date )4'1 Dig ita ll y signed by Gregory 5. Ki nn , DN: cn=Gregory
                                                                                          ' -' - -
: 5. Kinn, o=Wo lf C reek , /J; ' ou=5 up erviso r Reactor En g in ee rin g/C D/Fuel, * -~ e mail=g r kinn@wcnoc.com, c=U5 Dat e: 20 1 6.10.1 9 0 1:53: 0 3 -05'00' Gregory S. Kinn Date DC12 10/26/2016 Page 1 of 16 
                                                                                                      ' -
* *
6 40                                                  -........_                               -' '
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 22 has been prepared in accordance with the requirements of Technical Specification 5.6.5. The core operating limits that are included in the COLR affect the following Technical Specifications
                                                                                                                            -- .
: 2.1.1 Reactor Core Safety Limits 3.1.1 Shutdown Margin (SOM) 3.1.3 Moderator Temperature Coefficient (MTC) 3. 1 .4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 PHYSICS TESTS Exceptions  
                        ----
-MODE 2 3.2.1 3.2.2 3.2.3 3.3.1 3.4.1 Heat Flux Hot Channel Factor (F 0 (z)) (F a Methodology)
                                                                                                                    '
Nuclear Entha l py Rise Hot Channel Factor (F!) AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology)
2 000 p s ia
Reactor Tr i p System (RTS) Instrumentation RCS Pressure, Temperature , and Flow Departure from Nucleate Boiling (DNB) Limits 3.9.1 Boron Concentrat i on The portions of the Technical Specif i cation Bases affected by the report are listed below: ASA B 3.4.1 RCS Pressure , Temperature , and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 16
                                                                                                                        ' '
* *
U:::-
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 '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. 680 660 640 'Z--C> > .,, I-Q) UJ UJ 620 Q) > Q) :c ca 3:: ..2 cl: 600 580 560 0.0 I 24 00 p s i a ---_ I 225 0 p s i a A cce ptabl e Op e ration 0.2 0.4 0.6 0.8 Un acce ptabl e Op e ration " ' ' . . 1.0 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 16 ' ' 1.2 
    ':!..-                                      /                              ------- ~ ;--...,.
* *
                                                                  -- -----
* W$LFCREEK Wolf Creek Generating Station Cycle 22 'NUCLEAR OPERATING CORPORATION 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/&deg;F. The 300 PPM MTC Surveillance limit is -41 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
                        ........__                                                                           r---_
                                            ~
Cl
            >
                                        ~
                                                                                                ~
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i-:'
      .,,.,,                                        r--..                          225 0 p s ia
                                                                                                                                ~    ' *.
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      >
Q)      620 I'--..
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                                                                                                                                      \
                                                                                                                                          '  .',
                                                                  ------ -----
Q) 1925 p s ia
                                                ----                                   I'--                                             \.
                                                                                                                                            ' ' \. .
::0 cu
:3:                                                                                         r-----..                                        ''
      ..2
      <t:                                                                                      ----r----..       ----  ~
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                                                                                                                                                '
                                                                                                                                                    \
                                                                                                                                                        ''
                                                                                                                                                      \.
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/&deg;F.
The 300 PPM MTC Surveillance limit is -41 pcml&deg;F (equilibrium, all rods withdrawn , RATED THERMAL POWER condition) .
The 60 PPM MTC Surveillance limit is -46 pcm/&deg;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 &deg;/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 &deg;/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 &deg;(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/&deg;F coefficient Overtemperature     ~T reactor trip setpoint pressure      K3 = 0.000671/psig coefficient Nominal T avg at RTP                                        T'  ~ 586 .5&deg;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/&deg;F for increasing T avg rate/lag coefficient                                   = 0/&deg;F for decreasing T avg Overpower T reactor trip setpoint T avg heatup     K 6 = 0.00128/&deg;F for T > T" coefficient                                           = 0/&deg;F for T ~ T" Indicated T avg at RTP (calibration temperature   T"  ~ 586 .5&deg;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 &deg;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$LFCREEK  'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report
* Revision 0 2.2       Moderator Temperature Coefficient (MTC) (LCO 3.1 .3, SR 3.1.3.2)
The MTC shall be less positive than the limit provided in Figure 2.2.
The MTC shall be less negative than -50 pcm/&deg;F.
The 300 PPM MTC Surveillance limit is -41 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/&deg;F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
ii:' E g_ 8 ;: 6 z !!! C,) u: u. w 0 C,) w ai:: :, 4 ... cc ai:: w D,. :E w ... ai:: 0 ... CC 2 ai:: w Cl 0 :E 0 --0 1 0 20 I I UNACCEPT'}BLE bPERATUi)N 6.0 , 70% A r_CEPTAB1j-E OPERATION 30 40 50 60 70 80 90 % of RATBlTHERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs. THERMAL POWER (%) Page 4 of 16 100 
8 I      I UNACCEPT'}BLE bPERATUi)N ii:'
* *
      ~
* W~LFCREEK 'NUCLEAR OPERAT I NG CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 2.3 Shutdown Bank Insertion Limits (LCO 3.1.5) The shutdown banks shall be fully withdrawn (i.e., positioned within the interval of 222 and 231 steps withdrawn).
E                                                         6 .0 , 70%
2.4 Control Bank Insertion Limits (LCO 3.1.6) The Control Bank insertion , sequence, and overlap limits are specified in Figure 2.4. (FULLY WITHDRAWN) 220 200 .70~
g_
* t) I 180 ( 100% . 1 1 ) s T E 140 p s W 120 I T H 100 D R A 80 w N 60 40 20 0 ( 30 2o/o 0) 0 20 40 60 80 (FULLY INSERTED)
      ;: 6 z
THERMAL POWER (Percent)
      !!!
Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs. THERMAL POWER (%) -Four Loop Operation Fully withdrawn shall be the condition where control banks are at a position within the interval of~ 222 and :5: 231 steps withdrawn. Page 5 of 16 100 
C,)
* *
u:
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 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 O/o 0 F 90 R A T E D80 T H E R70 M A L P50 0 w E R 50 40 -40 UNACCEPTABLE OPERATION ( -29 , 50) -30 ( -15 , 100 ) (5 , 100) UNACCEPTABLE OPERATION ACCEPTABLE OPERATION ( 24 , 50) 10 0 10 20 AXIAL FLUX DIFFERENCE
u.
(%A I) Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER (%) Page 6 of 16 30 40 
w
* *
* 0 C,)
* W$LFCREEK  
w ai::
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (Fo(Z J){F a Methodology) (LCO 3.2.1, SR 3.2.1.2) FQ(Z) CF Q *K(Z), f o r P > 0.5 p FQ(Z) c;; *K(Z), for P 0.5 where , P = CFQ = pR TP = Q THERMAL P O WE R RATE D THERMAL PO WE R pRTP Q F Q(Z) limit at RATED THERMAL POWER (RTP) = 2.50, and K(Z) = as defined in Figure 2.6. F QM (Z) i s the measured value of F Q(Z), inferred from a power distribution measurement obtained with the Movable lncore Detector System (MIDS) or the Power Distribution Monitoring System. Measurement uncertainty is applied as follows. F J(Z)=F QM(Z)(1.03)(1.
      ...cc
0 5)=F t(Z)(1.08 I 5) when F QM(Z) i s obta i ned from MIDS. F J (Z) = F QM (Z)(l.03 )(U Qu) when F QM (Z) is obtained from PDMS. Manufacturing tolerances are accounted for in the 1.03 Engineering uncertainty factor. Measurement uncertainty for MIDS is accounted for in the 1.05 factor. PDMS measurement uncertainty is accounted for in the Uau factor , and it is determined by PDMS. F; (Z)=FJ (Z) W(Z) where, W(Z) = a cycle dependent function that accounts for power distribution transients encountered during normal operation (see Appendix A). When using the PDMS , F; (Z) uses F J (Z) that is determined from an F QM (Z) that reflects full-power steady-state conditions rather than current conditions . See Appendix A for: FQ Penalty Factor. Page 7 of 16
:,    4 ai::                       A r_CEPTAB1j-E w
* *
D,.                         OPERATION
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 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
:E
* *
      ...
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION 2.7 Nuclear Enthalpy Rise Hot Channel Factor ( F;) (LCO 3.2.2) Fi shall be limited by the following relationship:
w ai::
Fi ~Fir[1.0+PFMl(l.O-P)] Where , F:;p = Fi limit at RA TED THERMAL POWER (RTP) = 1.650 PF Ml = power factor multiplier for F i p = 0.3 = THE R MAL POWE R RATED THE R MAL PO WER = F i is the measured value of Fi, inferred from a power distribution measurement obta i ned with the Movable lncore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS). Measurement uncertainty is applied as follows. When Fi is obtained from MIDS , the measured value is multiplied by 1.04. When Fi is obtained from PDMS, the measured value is increased by an uncertainty factor (U~H), and the factor is determined by PDMS , with a lower limit of 4% . Page 9 of 16
      ...
* *
0 CC     2   -    -
* W$LFCREEK Wolf Creek Generating Station Cycle 22 '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 Overtemperature T reactor trip setpoint Overtemperature T reactor trip setpoint T avg coefficient Overtemperature T reactor trip setpoint pressure coefficient Nominal T avg 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 average temperature lead/lag constant Value K1 = 1.10 K2 = 0.0137/&deg;F K3 = 0.000671/psig T' 586.5&deg;F P' 2 2235 psig 11 = 6 sec 12 = 3 sec 13 = 2 sec 14 = 16 sec 15 = 4 sec 15 = O sec 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
ai::
* *
w Cl 0
* W$LFCREEK Wolf Creek Generating Station Cycle 22 'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower~ T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1 , Note 2) Parameter Overpower~ T reactor trip setpoint Overpower~ T reactor trip setpoint T avg rate/lag coefficient Overpower T reactor trip setpoint T avg heatup coefficient Indicated Tavg at RTP (calibration temperature for~ T instrumentation)
:E 0
Measured RCS T lead/lag constant Measured RCS T lag constant Measured RCS average temperature lead/lag constant Measured RCS average temperature rate/lag constant h (~I) = 0% RTP for all ~I Page 11 of 16 Value K 4=1.10 Ks= 0.02/&deg;F for increasing Tavg = 0/&deg;F for decreasing T avg Ke= 0.00128/&deg;F for T > T" = 0/&deg;F for T .:=; T" T" .:=; 586.5&deg;F 1"1 = 6 sec 1"2 = 3 sec 1"3 = 2 sec 1"e = O sec 1" 1 = 1 O sec
0     10    20     30        40      50    60      70      80      90  100
* *
                                          % of RATBlTHERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs.
* W$LFCREEK  
THERMAL POWER (%)
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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 &deg;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 1. 76 WRB-2 Design Limit DNBR 1.23 Page 12 of 16
* Page 4 of 16
* * * . '* W$LFCREEK
 
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 A. APPENDIX A Input relating to LCO 3.2.1: F ( Z) m ax tran s i ent l W ( Z) = Q x -for P > 0.5 FQ ( zt ea d y s tate p, F ( Z) m ax tr a n s ient l W(Z) = Q x-, for P 0.5 FQ(zteact ys t a te 0_5 where , THERMAL POWER P= RA TED THERMAL P OWER F Q(Zl'a t tr ans i e nt = 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(zf e ad ys t a t e = (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 st a te (denominator in above equations) is generated at the specific anticipated surveillance conditions.
Wolf Creek Generating Station W~ LFCREEK
W(Z) values are issued in controlled reports which will be provided on request. Page 13 of 16
          'NUCLEAR OPERATING CORPORATION Cycle 22 Core Operating Limits Report
* *
* Revision 0 2.3   Shutdown Bank Insertion Limits (LCO 3.1.5)
* W$LFCREEK Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 'NUCLEAR OPERATING CORPORATION Input relating to SR 3.2.1.2 Cycle Burnup F Q (Z) Penalty Factor (MWD/MTU) O to :5 7861 8059 8257 8454 8652 8850 9047 9245 Cycle Burnup (MWD/MTU)
The shutdown banks shall be fully withdrawn (i.e., positioned within the interval of ~ 222 and ~ 231 steps withdrawn).
:5 8,000 > 8,000 (%) 2.00 2.01 2.24 2.43 2.53 2.36 2.15 2.00 F Q (Z) Exclusion Zone (% [INCORE mesh points]) Top 15 [11] 10 [7] Page 14 of 16 Bottom 15 [11] 10 [7] 
2.4   Control Bank Insertion Limits (LCO 3.1.6)
* *
The Control Bank insertion, sequence, and overlap limits are specified in Figure 2.4.
* weLFCREEK  
(FULLY WITHDRAWN) 220 200
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station 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).
                                                                                              . 70~
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 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 -Fa Surveillance Technical Specification
* t) 180 I
," 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).
( 100% . 1 1 )
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
s T
* *
E 140
* W!LFCREEK  
* p s
'NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Cycle 22 Core Operating Limits Report Revision 0 7. WCAP-16009-P-A , "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncerta i nty Method (ASTRUM)," Revision 0 , January 2005. NRC letter dated November 5 , 2004 ," Final Safety Evaluation for WCAP-16009-P , Revis i on 0 , " Realistic La r ge 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
W   120 I
-P , Revision 0 , "Qualification of the 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)
T H 100 D
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+
R A   80 w
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 i x B, Addendum 1, 'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M86416)." 12. WCAP-12610-P-A  
N 60 40 20
& CENPD-404-P-A , Addendum 1-A , " Optimized Zirlo&#x17d;," 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&#x17d;," (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  
( 30 2o/o 0) 0 0                20                   40               60                 80                 100 (FULLY INSERTED)                       THERMAL POWER (Percent)
~T and Thermal Overtemperature  
Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs.
~T Trip Functions."' Page 16 of 16}}
* 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/&deg;F coefficient Overtemperature ~ T reactor trip setpoint pressure         K3 = 0.000671/psig coefficient Nominal T avg at RTP                                       T' ~ 586.5&deg;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/&deg;F for increasing Tavg rate/lag coefficient                                 = 0/&deg;F for decreasing T avg Overpower ~ T reactor trip setpoint Tavg heatup   Ke= 0.00128/&deg;F for T > T" coefficient                                          = 0/&deg;F for T .:=; T" Indicated Tavg at RTP (calibration temperature   T" .:=; 586.5&deg;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 &deg;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}}

Revision as of 05:35, 21 October 2019

Submittal of Cycle 22 and Cycle 23 Core Operating Limits Report
ML18127A072
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
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


Text

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

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

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

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
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