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Category:Fuel Cycle Reload Report
MONTHYEARML24109A1842024-04-18018 April 2024 Cycle 27 Core Operating Limits Report ML22303A0012022-10-30030 October 2022 Cycle 26 Core Operating Limits Report ML22110A2152022-04-20020 April 2022 Cycle 25 Core Operating Limits Report, Revision 1 ML21117A4212021-04-27027 April 2021 Cycle 25 Core Operating Limits Report ML19302D5592019-10-22022 October 2019 Submittal of Cycle 24 Core Operating Limits Reports ML18127A0722018-04-29029 April 2018 Submittal of Cycle 22 and Cycle 23 Core Operating Limits Report ML15314A6582015-11-0404 November 2015 Cycle 21 Core Operating Limits Report, Revision 1 ML15098A1162015-03-31031 March 2015 Cycle 21 Core Operating Limits Report ET 09-0001, Cycle 17 Core Operating Limits Report, Revision 12009-01-0707 January 2009 Cycle 17 Core Operating Limits Report, Revision 1 ET 08-0031, Cycle 17 Core Operating Limits Report2008-05-10010 May 2008 Cycle 17 Core Operating Limits Report ET 06-0051, Generation Station Cycle 16 Core Operating Limits Report2006-11-14014 November 2006 Generation Station Cycle 16 Core Operating Limits Report ET 05-0005, Cycle 15 Core Operating Limits Report2005-05-0505 May 2005 Cycle 15 Core Operating Limits Report WO 04-0010, Transmittal of Cycle 14 Core Operating Limits Report2004-02-23023 February 2004 Transmittal of Cycle 14 Core Operating Limits Report ET 02-0020, Rev. 0, Cycle 13 Core Operating Limits Report2002-04-30030 April 2002 Rev. 0, Cycle 13 Core Operating Limits Report 2024-04-18
[Table view] Category:Letter type:ET
MONTHYEARET 23-0006, CFR 50.55a Request Number CI3R-01 for the Third Containment Inservice Inspection Program Interval for Proposed Alternative Frequency to Containment Unbonded Post-Tensioning System Components2023-05-17017 May 2023 CFR 50.55a Request Number CI3R-01 for the Third Containment Inservice Inspection Program Interval for Proposed Alternative Frequency to Containment Unbonded Post-Tensioning System Components ET 23-0005, 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peen2023-03-16016 March 2023 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened ET 23-0003, License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications2023-03-0101 March 2023 License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications ET 23-0002, Supplement to License Amendment Request to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections2023-02-0707 February 2023 Supplement to License Amendment Request to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ET 23-0004, Response to Requests for Additional Information (RAI) Regarding License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements2023-01-26026 January 2023 Response to Requests for Additional Information (RAI) Regarding License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements ET 22-0006, Operating Corp. - Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections2022-12-0101 December 2022 Operating Corp. - Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections ET 22-0010, License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements2022-08-0202 August 2022 License Amendment Request (LAR) for Deviation from Fire Protection Program Requirements ET 22-0012, Supplement to License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating2022-07-12012 July 2022 Supplement to License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ET 22-0011, Withdrawal of 10 CFR 50.55a Request I4R-08, Relief for Extension of Follow-up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface2022-05-31031 May 2022 Withdrawal of 10 CFR 50.55a Request I4R-08, Relief for Extension of Follow-up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface ET 22-0005, Response to Request for Additional Information Regarding License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating2022-04-13013 April 2022 Response to Request for Additional Information Regarding License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ET 22-0003, CFR 50.55a Request I4R-09 for the Fourth Inservice Inspection Program Interval, Relief from Examination of Reactor Vessel Flange Threads2022-04-0606 April 2022 CFR 50.55a Request I4R-09 for the Fourth Inservice Inspection Program Interval, Relief from Examination of Reactor Vessel Flange Threads ET 22-0002, Operating Corp 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head .2022-04-0404 April 2022 Operating Corp 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination and Visual Examination Requirements for Reactor Pressure Vessel Head . ET 22-0004, Operating Corp - Response to Request for Confirmation of Information (RCI) Regarding Steam Generator Tube Inspection Report2022-03-15015 March 2022 Operating Corp - Response to Request for Confirmation of Information (RCI) Regarding Steam Generator Tube Inspection Report ET 22-0001, Removal of the Table of Contents from the Technical Specifications2022-01-12012 January 2022 Removal of the Table of Contents from the Technical Specifications ET 21-0017, Response to Request for Additional Information Regarding License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation2021-12-22022 December 2021 Response to Request for Additional Information Regarding License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation ET 21-0009, Results of the Steam Generator Tube Inservice Inspection During the 24th Refueling Outage2021-11-0101 November 2021 Results of the Steam Generator Tube Inservice Inspection During the 24th Refueling Outage ET 21-0015, Withdrawal of License Amendment Request for a Risk-Informed Resolution to GSI-1912021-10-20020 October 2021 Withdrawal of License Amendment Request for a Risk-Informed Resolution to GSI-191 ET 21-0012, Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-1912021-10-11011 October 2021 Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-191 ET 21-0004, License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating2021-09-29029 September 2021 License Amendment Request - Diesel Generator Completion Time Extension for Technical Specification 3.8.1, AC Sources - Operating ET 21-0010, License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation2021-09-29029 September 2021 License Amendment Request - Revision to Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation ET 21-0005, Operating Corp., License Amendment Request for a Risk-Informed Resolution to GSI-1912021-08-12012 August 2021 Operating Corp., License Amendment Request for a Risk-Informed Resolution to GSI-191 ET 20-0013, Response to Request for Additional Information Re Application for Technical Specification Change Re Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program (TSTF-415)2020-10-26026 October 2020 Response to Request for Additional Information Re Application for Technical Specification Change Re Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program (TSTF-415) ET 20-0011, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI2020-10-0101 October 2020 Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI ET 20-0007, License Amendment Request for Replacement of Engineered Safety Features Transformers with New Transformers That Have Active Automatic Load Tap Changers2020-06-0808 June 2020 License Amendment Request for Replacement of Engineered Safety Features Transformers with New Transformers That Have Active Automatic Load Tap Changers ET 20-0008, Operating Corporation Update for Full Implementation of Open Phase Detection System2020-05-20020 May 2020 Operating Corporation Update for Full Implementation of Open Phase Detection System ET 20-0004, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425)2020-04-27027 April 2020 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425) ET 20-0002, Response to Request for Additional Information Regarding Utilizing Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 12020-01-29029 January 2020 Response to Request for Additional Information Regarding Utilizing Code Case N-666-1, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 1 ET 19-0021, Errata for Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation2019-12-0909 December 2019 Errata for Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ET 19-0020, Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation2019-11-13013 November 2019 Supplemental Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ET 19-0019, Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.2019-09-10010 September 2019 Response to Request for Additional Information Revision to Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. ET 19-0018, Supplement to License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.2019-08-22022 August 2019 Supplement to License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. ET 19-0014, lnserv1ce Inspection (ISI) Program Relief Request Number 14R-07, to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, D1v1s1on 12019-08-15015 August 2019 lnserv1ce Inspection (ISI) Program Relief Request Number 14R-07, to Utilize Code Case N-513-4, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, D1v1s1on 1 ET 19-0002, License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.2019-03-18018 March 2019 License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. ET 19-0008, Response to Request for Additional Information Related to Thermal Conductivity Degradation for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption.2019-03-0505 March 2019 Response to Request for Additional Information Related to Thermal Conductivity Degradation for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption. ET 19-0003, License Amendment Request to Revise Technical Specification 3.6.3 and Surveillance Requirement 3.6.3.1 to Remove Use of a Blind Flange2019-02-25025 February 2019 License Amendment Request to Revise Technical Specification 3.6.3 and Surveillance Requirement 3.6.3.1 to Remove Use of a Blind Flange ET 19-0001, Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 42019-01-23023 January 2019 Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules, Revision 4 ET 18-0032, Operating Corporation Change of Date for Full Implementation of Open Phase Detection System2018-12-0707 December 2018 Operating Corporation Change of Date for Full Implementation of Open Phase Detection System ET 18-0035, Operating Corp., Supplemental Response to RAI for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-12-0606 December 2018 Operating Corp., Supplemental Response to RAI for License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 18-0029, (WCGS) - Results of the Twenty First Steam Generator Tube Inservice Inspection2018-10-31031 October 2018 (WCGS) - Results of the Twenty First Steam Generator Tube Inservice Inspection ET 18-0018, Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-06-19019 June 2018 Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 18-0016, Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools Request for Supplemental Information2018-05-29029 May 2018 Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools Request for Supplemental Information ET 18-0014, Response to Request for Additional Information Regarding the License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System.2018-05-29029 May 2018 Response to Request for Additional Information Regarding the License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System. ET 18-0013, Relief Request Number I4R-06, Request for Relief from ASME Code Case N-729-4 for Reactor Vessel Head Penetration Nozzle Weld2018-05-0202 May 2018 Relief Request Number I4R-06, Request for Relief from ASME Code Case N-729-4 for Reactor Vessel Head Penetration Nozzle Weld ET 18-0011, (WCGS) - Guarantee of Payment of Deferred Premiums2018-04-30030 April 2018 (WCGS) - Guarantee of Payment of Deferred Premiums ET 18-0012, Operating Corp., Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-04-19019 April 2018 Operating Corp., Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 18-0010, Financial Protection Levels2018-03-29029 March 2018 Financial Protection Levels ET 18-0007, Supplement to License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System.2018-02-15015 February 2018 Supplement to License Amendment Request for Addition of New Technical Specification 3.7.20, Class 1E Electrical Equipment Air Conditioning (A/C) System. ET 18-0005, Withdrawal of License Amendment Request for Revision to the Emergency Plan2018-02-0505 February 2018 Withdrawal of License Amendment Request for Revision to the Emergency Plan ET 18-0004, Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term2018-01-29029 January 2018 Supplement to License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analysis Including Adoption of Alternative Source Term ET 17-0025, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term2017-11-14014 November 2017 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications to Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term 2023-05-17
[Table view] |
Text
UCLEAR OPERATING CORPORATION Terry J. Garrett May 10, 2008 Vice President, Engineering ET 08-0031 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Docket No. 50-482: Wolf Creek Generating Station Cycle 17 Core Operating Limits Report Gentlemen:
Enclosed is Revision 0 of the Wolf Creek Generating Station Cycle 17 Core Operating Limits Report (COLR). This document is being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station Technical Specifications.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4084, or Mr. Richard D. Flannigan at (620) 364-4117.
Sincerely, Terry J. Garrett TJG/rlt Enclosure cc: E. E. Collins (NRC), w/e V. G. Gaddy (NRC), w/e B. K. Singal (NRC), w/e Senior Resident Inspector (NRC), wle 4OcD (
P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET
Enclosure to ET 08-0031 Wolf Creek Generating Station Cycle 17 - Core Operating Limits Report, Rev. 0
Wolf Creek Generating Station W*LF CREEK
'NUCLEARI OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 17 CORE OPERATING LIMITS REPORT Revision 0 March 2008 Prepared by: 02/14/2008 Matthew K. Morris Date o:;ý Reviewed by: 3/3/2008 Jeff T. Blair Date Approved by: 03/04/08 William S. Kennamore Date Page 1 of 15
Wý WCLFR LF CREEKCycle 17 Station Wolf Creek Generating CR EG CCore
'NUCLEAR OPERATING CORPORATION Operating Limits Report Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 17 has been prepared in accordance with the requirements of Technical Specification 5.6.5.
The core operating limits that are included in the COLR affect the following Technical Specifications:
2.1.1 Reactor Core Safety Limits 3.1.1 Shutdown Margin (SDM) 3.1.3 Moderator Temperature Coefficient (MTC) 3.1.4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 PHYSICS TESTS Exceptions - MODE 2 3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (IL) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) 3.3.1 Reactor Trip System (RTS) Instrumentation 3.3.2 Engineered Safety Feature Actuation System (ESFAS)
Instrumentation 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.9.1 Boron Concentration The portions of the Technical Specification Bases affected by the report are listed below:
ASA B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Page 2 of 15
Wolf Creek Generating Station W;0LF CREEK FNUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the subsections below:
2.1 Reactor Core Safety Limits (SL 2.1.1)
In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits in Figure 2.1.
680 660 640
- 4) 620 600 580 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1 Reactor Core Safety Limits Page 3 of 15
Wolf Creek Generating Station W@LF INUCLEAR CREEK Cycle 17 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/°F.
The 300 PPM MTC Surveillance limit is -41 pcm/°F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/OF (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
8 UNACCEPTABLE OPERATION E 6.0, 70%
z L6 U'
0 ACCEPTABLE OPERATION I.-
19 0
0 0 10 20 30 40 50 60 70 80 90 100
% of RATED THERMAL POWER Figure 2.2 Moderator Temperature Coefficient Vs.
RATED THERMAL POWER Page 4 of 15
Wolf Creek Generating Station W LF CREEK
'NUCLEAR OPERATING CORPORATION Core Operating Limits Report Cycle 17 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 , 2 -ro -,222) i 76.7% 222 )
I . I I 200 T - - - - .- -. - II- - - - -
B I T - --..
II - r I I I I 180
I I
I I
I I I I
I - - - - %! 161 )
160 IF I
I II II I I I -R -
S T
E 140
- I-. -. .- - . -
P S I I I I I I I I I III I I I I F I 120 - -- -- - - - - - - - - - - - - - - - - - - - - - - -
W I
T III I II I H 100 D
R A 80 W
I N I- - - - - -
T FT------
-I
F 60 . . . J . . . . . I. . . . . .I _ L _ . . .I . . . . . JF. .
I (0%,4'6 40 I I I I I I I 20 0
0 20 40 60 80 100 (FULLY INSERTED) RATED THERMAL POWER (P e rc e n t)
Figure 2.4 Control Bank Insertion, Sequence, and Overlap Limits Vs.
THERMAL POWER - Four Loop Operation Fully withdrawn shall be the condition where control banks are at a position within the interval of - 222 and < 231 steps withdrawn.
Page 5 of 15
Wolf Creek Generating Station WCREEK NUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)
Methodology) (LCO 3.2.3)
The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space is defined by Figure 2.5.
110
(-15,100) (5,100) 100 UNACCEPTABLE UNACCEPTABLE
% OPERATION OPERATION 0
F 90 R
A T
E D 80 T
ACCEPTABLE H OPERATION E
R 70 M
A L
P 60 0
W E
R 50
-29 60 (-29 ,50(24 .50 40
-40 -30 -20 -10 0 10 20 30 40 AXIAL FLUX DIFFERENCE (%Al)
Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of RATED THERMAL POWER Page 6 of 15
Wolf Creek Generating Station W6LF CREEK FNUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (Fo(Z))(FQ Methodology) (LCO 3.2.1, SR 3.2.1.2)
FQ(Z)< -FQ*K(Z), for P > 0.5 P
FQ(Z)< CFQ*K(Z), forP
- 0.5 0.5 THERMAL POWER where, P RATED THERMAL POWER CFQ = FQrP FRTP = FQ(Z) limit at RATED THERMAL POWER (RTP)
= 2.50, and K(Z) = as defined in Figure 2.6.
F§ (Z) = FQj (Z)(1.03)(1.05) = FQ" (Z)(1.0815) where, FQj(Z) = Measured value of FQ(Z) from incore flux map Fw (Z)=F, (Z) W(Z) where, W(Z) = a cycle dependent function that accounts for power distribution transients encountered during normal operation (see Appendix A).
See Appendix A for:
Fo Penalty Factor Page 7 of 15
Wolf Creek Generating Station W*LF CREEK
'NUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 1.2 N 1.0 -~----------I------
0 C.)
L- 0.8 Z
U-0.6 0w 0
zU 0.4 F RIP' 2.50 0
Z 0.2 ---------- Elevation (ft) K(Z) 0.0 1.000 6.0 1.000
'1 12. 0.925 1 0.0 0 2 4 6 8 10 12 CORE HEIGHT (FT)
Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Height Page 8 of 15
Wolf Creek Generating Station W4 LF CREEK
'NUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F-) (LCO 3.2.2)
F[ shall be limited by the following relationship:
FN <FTP[1.0 + PFv(1.0 - P)]
Where, FP = FA limit at RATED THERMAL POWER (RTP)
= 1.586 PFw = power factor multiplier for FA
= 0.3 P = THERMAL POWER RATED THERMAL POWER F, = Measured values of F; obtained by using the movable incore detectors to obtain a power distribution map. The measured values of F, shall be used since an uncertainty of 4% for incore measurement of F; has been included in the above limit.
Page 9 of 15
Wolf Creek Generating Station WuRLF REEK INUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.8 Reactor Trip System Overtemperature AT Setpoint Parameter Values (LCO 3.3.1)
Parameter Value Overtemperature AT reactor trip setpoint K= 1.10 Overtemperature AT reactor trip setpoint Tavg K2 = 0.0137/ 0F coefficient Overtemperature AT reactor trip setpoint pressure K3 = 0.000671/psig coefficient Nominal Tavg at RTP T' _ 586.50 F Nominal RCS operating pressure P' > 2235 psig Measured RCS AT lead/lag constant , =6 sec T2 = 3 sec Measured RCS AT lag constant T3= 2 sec Measured RCS average temperature lead/lag T4 = 16 sec constant T5 =4 sec Measured RCS average temperature lead/lag T6 =0 sec constant fl(Al) = -0.0227 {23% + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP < (qt-qb) < 5% RTP 0.0184 {(qt-qb) - 5%} when (qt-qb) > 5% RTP Where, qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
Page 10 of 15
Wolf Creek Generating Station W4LF FNUCLEAR CREEK OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.2)
Parameter Value Overpower AT reactor trip setpoint K4 = 1.10 Overpower AT reactor trip setpoint Tavg Ks5= 0.02/°F for increasing Tavg rate/lag coefficient = 0/'F for decreasing Tavg Overpower AT reactor trip setpoint Tavg heatup K6= 0.00128/°F forT> T" coefficient = 0/F for T!<T" Indicated Tavg at RTP (calibration temperature T" < 586.5 0 F for AT instrumentation)
Measured RCS AT lead/lag constant T, =6 sec T2 = 3 sec Measured RCS AT lag constant T3= 2 sec Measured RCS average temperature lead/lag T6 =0 sec constant Measured RCS average temperature rate/lag T7 = 10 sec constant f 2(AI) = 0% RTP for all Al Page 11 of 15
4 Wolf Creek Generating Station WLF CREEK
?NUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 2.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits (LCO 3.4.1)
Parameter Indicated Value "
Pressurizer pressure Pressure > 2220 psig RCS average temperature Tavg -<590.5 OF 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% Ak/k).
2.13 Departure from Nucleate Boiling Ratio (DNBR) Limits (B 3.4.1, ASA)
Safety Analysis DNBR Limit 1.76 WRB-2 Design Limit DNBR 1.23 Page 12 of 15
Wolf Creek Generating Station W*L,.F CREEK
'NUCLEAR OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 APPENDIX A A. Input relating to LCO 3.2.1:
F0 (Z)m- transient W(Z) = Ft(Z),,ody,,tate These values are issued in a controlled report which will be provided on request.
Input relating to SR 3.2.1.2 Cycle Burnup F.j (Z) Penalty Factor (MWD/MTU) (%)
_>0 2.00 Fw(Z) Exclusion Zone Cycle Burnup (%)
(MWD/MTU) Top Bottom
>0 15 15 Page. 13 of 15
Wolf Creek Generating Station W FNUCLEAR LF CREEK OPERATING CORPORATION Cycle 17 Core Operating Limits Report Revision 0 B. Approved Analytical Methods for Determining Core Operating Limits The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents. V
- 1. WCNOC Topical Report TR 90-0025 W01, "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station." (ET 90-0140, ET 92-0103)
NRC Safety Evaluation Report dated October 29, 1992, for the "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station."
- 2. WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.
NRC Safety Evaluation Report dated January 17, 1989, for the "Acceptance for Referencing of Licensing Topical Report WCAP-1 1397, Revised Thermal Design Procedure."
- 3. WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the Wolf Creek Generating: Station" (ET-91-0026, ET 92-0142, WM 93-0010,;WM 93-0028).
NRC Safety Evaluation Report dated September 30, 1993, forthe "Transient Analysis Methodology for the Wolf Creek Generating Station."
EPRI Topical Report NP-7450(A), "RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," including NRC Safety Evaluation Report dated January 25, 2001, "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," (TAC No. MA4311)." RETRAN-3D code is only utilized in the RETRAN-02 mode.
- 4. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.
NRC Safety Evaluation Report dated November 26, 1993, "Acceptance for Referencing of Revised Version of Licensing Topical Report WCAP-10216-P, Rev. 1, Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification" (TAC No. M88206).
- 5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017).
NRC Safety Evaluation Report dated March 10, 1993, for the "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."
- 6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054).
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Wolf Creek Generating Station W*I-,LF CREEK
'NUCLEAR OPERATING CORPORATION Core OprtngLmtsRpr e
Core Operating Limits Report Revisio n 0
- 7. WCAP-10266-P-A, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.
NRC letter dated November 13, 1986, "Acceptance for Referencing of Licensing Topical Report WCAP-1 0266 "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code.""
WCAP-10266-P-A, Addendum 1, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code Addendum 1:
Power Shape Sensitivity Studies," December 1987.
NRC letter dated September 15, 1987, "Acceptance for Referencing of Addendum 1 to WCAP-1 0266, BASH Power Shape Sensitivity Studies."
WCAP-10266-P-A, Addendum 2, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code Addendum 2:
BASH Methodology Improvements and Reliability Enhancements," May 1988 NRC letter dated January 20, 1988, "Acceptance for Referencing Topical Report Addendum 2 to WCAP-1 0266, Revision 2, "BASH Methodology Improvements and Reliability Enhancements."
- 8. WCAP-1 1596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.
NRC Safety Evaluation Report dated May 17, 1988, "Acceptance for Referencing of Westinghouse Topical Report WCAP-1 1596 - Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."
- 9. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"
September 1988.
NRC letter dated June 23, 1986, "Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP."
- 10. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.
NRC Safety Evaluation Reports dated July 1, 1991, "Acceptance for Referencing of Topical Report WCAP-1 2610, 'VANTAGE+ Fuel Assembly Reference Core' Report' (TAC NO. 77258)."
NRC Safety Evaluation Report dated September 15, 1994, "Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1,
'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M86416)."
- 11. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Function." September 1986.
NRC Safety Evaluation Report dated April 17, 1986, "Acceptance for Referencing of Licensing Topical Report WCAP-8745(P)/8746(NP), 'Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions."'
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