ML21117A421
| ML21117A421 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/27/2021 |
| From: | Benham R Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA 21-0040 | |
| Download: ML21117A421 (19) | |
Text
Ron Benham Director Nuclear and Regulatory Affairs April 27, 2021 RA 21-0040 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Docket No. 50-482: Wolf Creek Generating Station Cycle 25 Core Operating Limits Report Commissioners and Staff:
These documents are being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station Technical Specifications.
Enclosure I is Revision 0 of the Wolf Creek Generating Station Cycle 25 Core Operating Limits Report applicable to all modes.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204.
Sincerely, Ron Benham RDB/rlt Enclosure I: Wolf Creek Generating Station Cycle 25 Core Operating Limits Report, Rev. 0 cc: S. S. Lee (NRC), w/e S. A. Morris (NRC), w/e N. OKeefe (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 l Burlington, KS 66839 l 620-364-8831
Enclosure I to RA 21-0040 WOLF CREEK GENERATING STATION CYCLE 25 CORE OPERATING LIMITS REPORT, REVISION 0 (18 pages including this page)
APF 05-013-01, REV. 04 Enclosure I TR-94-0015 WCNOC Cycle 25 Core Operating Limits Report (COLR)
Revision 0 ENGINEERING REVIEW:
DRAFTER: N/A CHECKER: N/A ENGINEER: See attached.
SUPERVISOR: 04/13/2021 ELECTRONIC APPROVAL
- 1. APPROVED-MFG. MAY PROCEED
- 2. NOT APPROVED--RESUBMIT FINAL DOCUMENT/DRAWING-MFG. MAY PROCEED YES NO
- 3. APPROVED INFORMATION NOT CONTROLLED UNDER DESIGN PROCESS
- 4. ACCEPTABLE-MAINTAIN AS RECORD (INFO. ONLY)
- 5. RESTRICTED FOR WOLF CREEK PLANNING ONLY-MFG. MAY PROCEED YES NO APPROVAL OF THIS DOCUMENT/DRAWING DOES NOT RELIEVE SUPPLIER/CONTRACTOR FROM FULL COMPLIANCE WITH CONTRACT, SPECIFICATIONS AND/OR PURCHASE ORDER REQUIREMENTS.
COMMENTS:
VETIP (AI 05C-001): This document does not contain design information that requires an engineering Change Package.
Safety Related NOTE: DO NOT RELEASE this document until directed by Nuclear Engineering.
This document is to be released during Refuel 24 after core offload and before core reload.
P.O.#: N/A VENDOR MANUAL:
PAGE: N/A CHANGE PACKAGE #: INCORPORATED CHANGE DOCUMENT(S):
N/A N/A REV. # DC RELEASED:
DigsigDSR 3 0.50 W29 DC7 04/13/2021 COMPONENT NUMBER(S) N/A COMPONENT NUMBERS ARE FOR INITIAL (REV, W01) DATA LINKING ONLY. ADDITIONAL COMPONENT LINKS ARE MADE IN DATABASE ONLY.
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 WOLF CREEK GENERATING STATION CYCLE 25 CORE OPERATING LIMITS REPORT Revision 0 April 2021 Prepared by: 4/8/2021 Ian Miller Date Reviewed by: 4/8/2021 Matthew Thomas Date Approved by: 4/13/2021 Chad Lisle Date Page 1 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 25 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 FQZ (FQ Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor FNH 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) 3.3.1 Reactor Trip System (RTS) Instrumentation 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.9.1 Boron Concentration Page 2 of 16
Wolf Creek Generating Station 1 Cycle 25 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.
2250 psia 2460 psia Unacceptable Consequences 1925 psia 2000 psia Acceptable Consequences Figure 2.1 Reactor Core Safety Limits Page 3 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 2.2 Moderator Temperature Coefficient (MTC) (LCO 3.1.3, SR 3.1.3.1, SR 3.1.3.2)
The MTC shall be less positive than the limit provided in Figure 2.2.
The MTC shall be less negative than -50 pcm/F.
The 300 PPM MTC Surveillance limit is -41 pcm/F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/F (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
8 UNACCEPT ABLE OPERAT ION MODERATOR TEMPERATURE COEFFICIENT (pcm/ F) o 6.0, 70%
6 4
ACCEPT ABL E OPERAT ION 2
0 0 10 20 30 40 50 60 70 80 90 100
% o f RAT ED T HERM AL POWER Figure 2.2 Moderator Temperature Coefficient Vs.
THERMAL POWER (%)
Page 4 of 16
Wolf Creek Generating Station 1 Cycle 25 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 limits are specified in Figure 2.4. The Control Bank withdrawal sequence is A-B-C-D. The insertion sequence is the reverse of the withdrawal sequence. The difference between each sequential Control Bank position is 115 steps when not fully inserted and not fully withdrawn.
(F U L L Y W IT H D R A W N )
220 ( 2 6 .7 % , 2 2 2 ) ( 7 6 .7 % , 2 2 2 )
200 BANK B
180
( 100% , 161 )
160 ( 0% , 161 )
S T
E 140 BANK P C S
W 120 I
T H 100 D
R A BANK 80 W D N
60
( 0% , 46 )
40 20
( 3 0 .2 % , 0 )
0 0 20 40 60 80 100 (F U L L Y IN S E R T E D ) T H E R M A L P O W E R (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 16
Wolf Creek Generating Station 1 Cycle 25 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
( -1 5 , 1 0 0 ) ( 5 , 100 )
100 U N A C C EP T A B L E U N A C C EP T A B L E O P E R A T IO N O P E R A T IO N O
F 90 R
A T
E D 80 T
A C C EP T A B L E H O P E R A T IO N E
R 70 M
A L
P 60 O
W E
R 50
( -2 9 , 5 0 ) ( 24 , 50 )
40
-4 0 -3 0 -2 0 -1 0 0 10 20 30 40 A X IA L F L U X D IF F E R E N C E ( % I)
Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER (%)
Page 6 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 2.6 Heat Flux Hot Channel Factor (FQ(Z))(FQ Methodology) (LCO 3.2.1, SR 3.2.1.1, SR 3.2.1.2)
CFQ FQ (Z) *K(Z), for P > 0.5 P
CFQ FQ (Z) *K(Z), for P 0.5 0.5 THERMAL POWER where, P =
RATED THERMAL POWER CFQ = FQRTP FQRTP = 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 Incore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS).
Measurement uncertainty is applied as follows.
FQC (Z) FQM (Z)( 1.03 )( 1.05 ) FQM (Z)( 1.0815 ) when FQM (Z) is obtained from MIDS.
FQC (Z) FQM (Z)( 1.03 )(U QU ) 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 UQU factor, and it is determined by PDMS.
FQW (Z) FQC (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, FQW (Z) uses FQC (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 1 Cycle 25 Core Operating Limits Report Revision 0 1.2 1.0 NORMALIZED PEAKING FACTOR K(Z) 0.8 0.6 0.4 F QRTP = 2.50 0.2 Ele vatio n (ft) K(Z )
0.0 1.000 6.0 1.000 12.0 0.925 0.0 0 2 4 6 8 10 12 C OR E H EIGH T (FT)
Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Height Page 8 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 N
2.7 Nuclear Enthalpy Rise Hot Channel Factor ( FH ) (LCO 3.2.2)
FNH shall be limited by the following relationship:
H 10 FNH FRTP . P
. PFH 10 Where, FRTP H
= FNH limit at RATED THERMAL POWER (RTP)
= 1.650 PFH = power factor multiplier for FNH
= 0.3 P = THERMAL PO WER RATED THER MAL POWER FNH = FNH is the measured value of FNH , inferred from a power distribution measurement obtained with the Movable Incore Detector System (MIDS) or the Power Distribution Monitoring System (PDMS). Measurement uncertainty is applied as follows.
When FNH is obtained from MIDS, the measured value is multiplied by 1.04.
When FNH is obtained from PDMS, the measured value is increased by an uncertainty factor (UH), and the factor is determined by PDMS, with a lower limit of 4%.
Page 9 of 16
Wolf Creek Generating Station 1 Cycle 25 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 Tavg K2 = 0.0137/F coefficient Overtemperature T reactor trip setpoint pressure K3 = 0.00095/psi coefficient Nominal Tavg (Tref from Rod Control) at RTP T 586.5F Nominal RCS operating pressure P 2235 psig Measured RCS T lead/lag constant 1 = 6 sec 2 = 3 sec Measured RCS T lag constant 3 = 2 sec Measured RCS average temperature lead/lag 4 = 16 sec constant 5 = 4 sec Measured RCS average temperature lead/lag 6 = 0 sec constant f1(I) = -0.0227 / %RTP {23% RTP + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP (qt-qb) 5% RTP 0.0184 / %RTP {(qt-qb) - 5% RTP} when (qt-qb) > 5% RTP Where, qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
Page 10 of 16
Wolf Creek Generating Station 1 Cycle 25 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 Tavg K5 = 0.02/F for increasing Tavg rate/lag coefficient = 0/F for decreasing Tavg Overpower T reactor trip setpoint Tavg heatup K6 = 0.00128/F for T T coefficient = 0/F for T T Nominal Tavg (Tref from Rod Control) at RTP T 586.5F Measured RCS T lead/lag constant 1 = 6 sec 2 = 3 sec Measured RCS T lag constant 3 = 2 sec Measured RCS average temperature lag 6 = 0 sec constant Measured RCS average temperature rate/lag 7 = 10 sec constant f2(I) = 0% RTP for all I Page 11 of 16
Wolf Creek Generating Station 1 Cycle 25 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 2219 psig (Average of 4 channels) 2221 psig (Average of 3 channels)
RCS average temperature Tavg 590.8 °F (Average of 4 channels) 590.6 °F (Average of 3 channels)
RCS total flow rate Flow 376,000 gpm 2.11 Boron Concentration (LCO 3.9.1)
The refueling boron concentration shall be greater than or equal to 2300 ppm.
2.12 SHUTDOWN MARGIN (LCO 3.1.1, 3.1.4, 3.1.5, 3.1.6, & 3.1.8)
The SHUTDOWN MARGIN shall be greater than or equal to 1300 pcm (1.3% k/k).
Page 12 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 APPENDIX A A. Input relating to LCO 3.2.1:
FQ ( Z ) max transient 1 W (Z ) steady state
, for P > 0.5 FQ ( Z ) P FQ ( Z ) max transient 1 W (Z ) , for P 0.5 FQ ( Z ) steady state 0.5 THERMAL POWER where, P=
RATED THERMAL POWER FQ(Z)max transient = Maximum (FQ(Z) x p) calculated over the entire range of power shapes analyzed for Condition I operations (p = power at which maximum occurs).
FQ(Z)steady state = (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 FQ(Z)steady state (denominator in above equations) is generated at the specific anticipated surveillance conditions.
W(Z) values are issued in controlled reports which will be provided on request.
Page 13 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 Input relating to SR 3.2.1.2 Cycle Burnup FQ ( Z ) Penalty Factor (MWD/MTU) (%)
0 to 348 2.00 546 2.70 743 3.21 941 3.33 1139 3.27 1337 3.17 1534 3.03 1732 2.84 1930 2.61 2128 2.35 2326 2.05 2523 2.00 8259 2.00 8457 2.49 8655 2.76 8853 2.96 9050 3.06 9248 3.14 9446 3.21 9644 3.26 9841 3.29 10039 2.00 FQ ( Z ) Exclusion Zone
(% [INCORE mesh points])
Cycle Burnup (MWD/MTU) Top Bottom 1000 5 [4] 5 [4]
> 1000 to < 8000 15 [11] 15 [11]
8000 10 [7] 10 [7]
Page 14 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0 B. Approved Analytical Methods for Determining Core Operating Limits The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.
- 1. WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989.
NRC Safety Evaluation Report dated January 17, 1989, for the Acceptance for Referencing of Licensing Topical Report WCAP-11397, Revised Thermal Design Procedure.
- 2. WCAP-10216-P-A, Revision 1A, Relaxation of Constant Axial Offset Control - 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).
- 3. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.
NRC Safety Evaluation Report dated May 28, 1985, Acceptance for Referencing of Licensing Topical Report WCAP-9272(P)/9273(NP), Westinghouse Reload Safety Evaluation Methodology.
- 4. WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM), Revision 0, January 2005.
NRC letter dated November 5, 2004,Final Safety Evaluation for WCAP-16009-P, Revision 0, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM) (TAC NO. MB9483).
- 5. WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004.
NRC Safety Evaluation dated March 18, 2004, Final Safety Evaluation for Westinghouse Topical Report WCAP-16045-P, Revision 0, Qualification of the Two-Dimensional Transport Code PARAGON.
- 6. WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology, August 2007.
NRC Safety Evaluation dated February 23, 2007, Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16045-P-A, Addendum 1, Qualification of the NEXUS Nuclear Data Methodology (TAC NO. MC9606).
- 7. WCAP 10965-P-A, ANC: A Westinghouse Advanced Nodal Computer Code, September 1986.
NRC letter dated June 23, 1986, Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP.
Page 15 of 16
Wolf Creek Generating Station 1 Cycle 25 Core Operating Limits Report Revision 0
- 8. WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995.
NRC Safety Evaluation Reports dated July 1, 1991, Acceptance for Referencing of Topical Report WCAP-12610, VANTAGE+ Fuel Assembly Reference Core Report (TAC NO. 77258).
NRC Safety Evaluation Report dated September 15, 1994, Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1, Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models (TAC NO.
M86416).
- 9. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZirloTM, July 2006.
NRC Safety Evaluation dated June 10, 2005, Final Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A, Optimized ZirloTM, (TAC NO. MB8041).
- 10. WCAP-8745-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