ML24337A013
| ML24337A013 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/02/2024 |
| From: | Hamman D Wolf Creek |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 000717 | |
| Download: ML24337A013 (1) | |
Text
P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Dustin T. Hamman Director Nuclear and Regulatory Affairs December 2, 2024 000717 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Docket No. 50-482: Wolf Creek Generating Station Cycle 27 Core Operating Limits Report, Revision 1 Commissioners and Staff:
The enclosed document is being submitted pursuant to Section 5.6.5 of the Wolf Creek Generating Station (WCGS) Technical Specifications.
Enclosure I is Revision 1 of the WCGS Cycle 27 Core Operating Limits Report (COLR) 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, Dustin T. Hamman DTH/jkt Enclosure I:
Wolf Creek Generating Station Cycle 27 Core Operating Limits Report, Revision 1 cc:
A. N. Agrawal (NRC), w/e S. S. Lee (NRC), w/e J. D. Monninger (NRC), w/e Senior Resident Inspector (NRC), w/e WC Licensing Correspondence - RA 24-000717, w/e
Enclosure I to 000717 WOLF CREEK GENERATING STATION CYCLE 27 CORE OPERATING LIMITS REPORT, REVISION 1 (17 pages)
WCNOC Cycle 27 Core Operating Limits Report (COLR)
Revision 1 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.
P.O.#: N/A VENDOR MANUAL:
PAGE: N/A CHANGE PACKAGE #:
N/A INCORPORATED CHANGE DOCUMENT(S):
N/A REV. #
DC RELEASED:
W33 DigsigDSR 3 0.50 COMPONENT NUMBER(S) N/A COMPONENT NUMBERS ARE FOR INITIAL (REV, W01) DATA LINKING ONLY. ADDITIONAL COMPONENT LINKS ARE MADE IN DATABASE ONLY.
ENGINEERING REVIEW:
DRAFTER: N/A CHECKER: N/A ENGINEER: See attached.
SUPERVISOR:
11/12/2024 Enclosure I (Page 1 of 17)
WOLF CREEK GENERATING STATION CYCLE 27 CORE OPERATING LIMITS REPORT Revision 1 November 2024 Prepared by:
10/24/2024 Ian Miller Date Reviewed by:
10/30/2024 Matthew Thomas Date Approved by:
11/12/2024 Chad Lisle Date Enclosure I (Page 2 of 17)
1 Page 2 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
1.0 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating Station Cycle 27 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
Z FQ (FQ Methodology) 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor
N H
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 Enclosure I (Page 3 of 17)
1 Page 3 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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.
Figure 2.1 Reactor Core Safety Limits Unacceptable Consequences 1925 psia 2460 psia 2250 psia 2000 psia Acceptable Consequences Enclosure I (Page 4 of 17)
1 Page 4 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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/qF.
The 300 PPM MTC Surveillance limit is -41 pcm/qF (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
The 60 PPM MTC Surveillance limit is -46 pcm/qF (equilibrium, all rods withdrawn, RATED THERMAL POWER condition).
UNACCEPTABLE OPERATION ACCEPTABLE OPERATION 6.0, 70%
0 2
4 6
8 0
10 20 30 40 50 60 70 80 90 100
% of RATED THERM AL POWER MODERATOR TEMPERATURE COEFFICIENT (pcm/
oF)
Figure 2.2 Moderator Temperature Coefficient Vs.
THERMAL POWER (%)
Enclosure I (Page 5 of 17)
1 Page 5 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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 )
( F U L L Y IN S E R T E D )
( 2 6.7 %, 2 2 2 )
( 0 %, 1 6 1 )
( 7 6.7 %, 2 2 2 )
( 0 %, 4 6 )
( 1 0 0 %, 1 6 1 )
( 3 0.2 %, 0 )
0 2 0 4 0 6 0 8 0 1 0 0 1 2 0 1 4 0 1 6 0 1 8 0 2 0 0 2 2 0 0
2 0 4 0 6 0 8 0 1 0 0 T H E R M A L P O W E R ( P e r c e n t )
S T
E P
S W
I T
H D
R A
W N
B A N K B
B A N K C
B A N K D
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 t 222 and d 231 steps withdrawn.
Enclosure I (Page 6 of 17)
1 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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.
( -1 5, 1 0 0 )
( 5, 1 0 0 )
( -2 9, 5 0 )
( 2 4, 5 0 )
4 0 5 0 6 0 7 0 8 0 9 0 1 0 0 1 1 0
-4 0
-3 0
-2 0
-1 0 0
1 0 2 0 3 0 4 0 AX IAL F L U X D IF F E R E N C E (% ' I)
O F
R A
T E
D T
H E
R M
A L
P O
W E
R U N A C C EP T A B L E O P ER A T IO N U N A C C EP T A B L E O P ER A T IO N A C C EP T A B L E O P ER A T IO N Figure 2.5 AXIAL FLUX DIFFERENCE Limits as a Function of THERMAL POWER (%)
Page 6 of 16 Enclosure I (Page 7 of 17)
1 Page 7 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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) 5 0.
or P >
- K(Z), f P
CFQ (Z)
FQ d
5 0
5 0
or P
- K(Z), f CFQ (Z)
FQ d
d where, P
=
POWER THERMAL RATED POWER THERMAL CFQ =
RTP Q
F RTP Q
F
=
(Z)
FQ limit at RATED THERMAL POWER (RTP)
= 2.50, and
Z K
= as defined in Figure 2.6.
(Z)
F M Q
is the measured value of (Z)
FQ
, 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.
)
(Z)(
F
)
)(
(Z)(
F (Z)
F M
Q M
Q C
Q 0815 1
05 1
03 1
when (Z)
F M Q
is obtained from MIDS.
)
)(U (Z)(
F (Z)
F QU M
Q C
Q 03 1
when (Z)
F M Q
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.
(Z)W(Z)
F (Z)
F C
Q W
Q where, W(Z) = a cycle dependent function that accounts for power distribution transients encountered during normal operation (see Appendix A).
When using the PDMS, (Z)
F W Q
uses (Z)
F C Q
that is determined from an (Z)
F M Q
that reflects full-power steady-state conditions rather than current conditions.
See Appendix A for:
Q F Penalty Factor.
Enclosure I (Page 8 of 17)
1 Page 8 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
0.0 0.2 0.4 0.6 0.8 1.0 1.2 0
2 4
6 8
10 12 CORE HEIGHT (FT)
NORMALIZED PEAKING FACTOR K(Z)
F Q RTP = 2.50 Ele vation (ft) K(Z) 0.0 1.000 6.0 1.000 12.0 0.925 Figure 2.6 K(Z) - Normalized Peaking Factor Vs. Core Height Enclosure I (Page 9 of 17)
1 Page 9 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
2.7 Nuclear Enthalpy Rise Hot Channel Factor (
N H
F' ) (LCO 3.2.2)
N H
F' shall be limited by the following relationship:
F F
PF P
H N
H RTP H
d
10 10 Where, F H RTP
= F H N
' limit at RATED THERMAL POWER (RTP)
= 1.650 PF H
= power factor multiplier for F H N
= 0.3 P
=
MAL POWER RATED THER WER THERMAL PO N
H F'
=
N H
F' is the measured value of N
H F', 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 N
H F' is obtained from MIDS, the measured value is multiplied by 1.04.
When N
H 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%.
Enclosure I (Page 10 of 17)
1 Page 10 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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 coefficient K2 = 0.0137/qF Overtemperature 'T reactor trip setpoint pressure coefficient K3 = 0.00095/psi Nominal Tavg (Tref from Rod Control) at RTP Tc d 586.5qF Nominal RCS operating pressure Pc t 2235 psig Measured RCS 'T lead/lag constant W1 = 6 sec W2 = 3 sec Measured RCS 'T lag constant W3 = 2 sec Measured RCS average temperature lead/lag constant W4 = 16 sec W5 = 4 sec Measured RCS average temperature lead/lag constant W6 = 0 sec f1('I) = -0.0227 / %RTP {23% RTP + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP d (qt-qb) d 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.
Enclosure I (Page 11 of 17)
1 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision 1 2.9 Reactor Trip System Overpower 'T Setpoint Parameter Values (LCO 3.3.1, Table 3.3.1-1, Note 2)
Parameter Value Overpower 'T reactor trip setpoint Overpower 'T reactor trip setpoint Tavg rate/lag coefficient Overpower 'T reactor trip setpoint Tavg heatup coefficient Nominal Tavg (Tref from Rod Control) at RTP Measured RCS 'T lead/lag constant Measured RCS 'T lag constant Measured RCS average temperature lag constant Measured RCS average temperature rate/lag constant K4 = 1.10 K5 = 0.02/qF for increasing Tavg
= 0/qF for decreasing Tavg K6 = 0.00128/qF for T ! Tcc
= 0/qF for T d Tcc Tcc d 586.5qF W1 = 6 sec W2 = 3 sec W3 = 2 sec W6 = 0 sec W7 = 10 sec f2('I) = 0% RTP for all 'I*
- Note: the f2(I) function generator has been physically removed by CP 012858; AFD is not an input to the OPDT trip setpoint. Reference AN-24-001-CN001.
Page 11 of 16 Enclosure I (Page 12 of 17)
1 Page 12 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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 d 590.8 qF (Average of 4 channels) d 590.6 qF (Average of 3 channels)
RCS total flow rate Flow t 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).
Enclosure I (Page 13 of 17)
1 Page 13 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
APPENDIX A A.
Input relating to LCO 3.2.1:
5 0
1
)
(
)
(
)
(
state steady ient max trans
, for P P
Z F
Z F
Z W
Q Q
u 5
0 5.0 1
)
(
)
(
)
(
state steady ient max trans
, for P Z
F Z
F Z
W Q
Q d
u
- where, P =
POWER THERMAL RATED POWER THERMAL 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.
Enclosure I (Page 14 of 17)
1 Page 14 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
Input relating to SR 3.2.1.2 Cycle Burnup (MWD/MTU)
)
(Z FQ Penalty Factor
(%)
0 to 3513 2.00 3711 2.33 3909 2.58 4106 2.66 4304 2.58 4502 2.21 4700 2.00 Cycle Burnup (MWD/MTU)
)
(Z FQ Exclusion Zone
(% [INCORE mesh points])
Top Bottom 3000 10 [7]
10 [7]
> 3000 to < 10000 15 [11]
15 [11]
10000 10 [7]
10 [7]
Enclosure I (Page 15 of 17)
1 Page 15 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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.
Enclosure I (Page 16 of 17)
1 Page 16 of 16 Wolf Creek Generating Station Cycle 27 Core Operating Limits Report Revision
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.
Enclosure I (Page 17 of 17)