ML17054C223
| ML17054C223 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/17/2017 |
| From: | Wolf Creek |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17054C103 | List: |
| References | |
| ET 17-0001 | |
| Download: ML17054C223 (8) | |
Text
1 Page 2 of 16 Wolf Creek Generating Station 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 (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 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 Attachment V to ET 17-0001 Page 2 of 9
1 Page 3 of 16 Wolf Creek Generating Station 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.
560 580 600 620 640 660 680 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Allowable Vessel Tavg (
oF)
Acceptable Operation Unacceptable Operation 2250 psia 2400 psia 1925 psia 2000 psia Figure 2.1 Reactor Core Safety Limits Replace with new core limits figure attached.
Attachment V to ET 17-0001 Page 3 of 9
Attachment V of ET 16-00XX Page 4 of 9 Unacceptable Consequences 1925 psia 2460 psia 2250 psia 2000 psia Acceptable Consequences Attachment V to ET 17-0001 Page 4 of 9
1 Page 10 of 16 Wolf Creek Generating Station 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 Tavg coefficient K2 = 0.0137/qF Overtemperature 'T reactor trip setpoint pressure coefficient K3 = 0.000671/psig Nominal Tavg 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 {23% + (qt-qb)} when (qt-qb) < -23% RTP 0% of RTP when -23% RTP d (qt-qb) d 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.
(Tref from Rod Control) 0.00095/psi
/ %RTP
/ %RTP RTP RTP Attachment V to ET 17-0001 Page 5 of 9
1 Page 11 of 16 Wolf Creek Generating Station 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 Tavg rate/lag coefficient K5 = 0.02/qF for increasing Tavg
= 0/qF for decreasing Tavg Overpower 'T reactor trip setpoint Tavg heatup coefficient K6 = 0.00128/qF for T ! Tcc
= 0/qF for T d Tcc Indicated Tavg at RTP (calibration temperature for 'T instrumentation)
Tcc d 586.5qF 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 W6 = 0 sec Measured RCS average temperature rate/lag constant W7 = 10 sec f2('I) = 0% RTP for all 'I Nominal (Tref from Rod Control)
Attachment V to ET 17-0001 Page 6 of 9
1 Page 12 of 16 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 t 2220 psig RCS average temperature Tavg d 590.5 qF RCS total flow rate Flow t 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 2219 (Average of 4 channels) 2221 (Average of 3 channels) 590.8°F (Average of 4 channels) 590.6°F (Average of 3 channels) 376,000 ppm Delete Section 2.13 Attachment V to ET 17-0001 Page 7 of 9
1 Page 15 of 16 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 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 - 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).
1.
2.
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.
3.
Attachment V to ET 17-0001 Page 8 of 9
1 Page 16 of 16 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 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.
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, 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).
- 12. 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).
- 13. 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.
4.
5.
6.
7.
8.
9.
10.
Attachment V to ET 17-0001 Page 9 of 9