L-PI-12-027, Core Operating Limits Report (COLR) for Prairie Island Nuclear Generating Plant (PINGP) Unit 2, Cvcle 27, Revision 0

Core Operating Limits Report (COLR) for Prairie Island Nuclear Generating Plant (PINGP) Unit 2, Cvcle 27, Revision 0

ML12109A307
Person / Time
Site:	Prairie Island
Issue date:	04/18/2012
From:	Schimmel M Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-12-027
Download: ML12109A307 (32)

Text

a I 1\

Xcel EnergyB APR 1 8 2012 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Unit 2 Docket 50-306 Renewed License No. DPR-60 Core Operating Limits Report (COLR) for Prairie Island Nuclear Generating Plant

[PINGP) Unit 2, Cvcle 27, Revision 0 Pursuant to the requirements of Technical Specification 5.6.5.d1Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"),

submits the COLR for the PINGP Unit 2, Cycle 27, Revision 0. The COLR provides the cycle-specific values of the limits established using NRC approved methodologies such that the applicable limits of the plant safety analysis are met.

The COLR for PINGP Unit 2, Cycle 27, Revision 0, is provided in Enclosure 1.

Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.

Mark A. Schimmel Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC State of Minnesota 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121

ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 2 - CYCLE 27 REVISION 0 Record of Revision (6 pages)

Unit 2 - Cycle 27, Revision 0 (24 pages)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Unit Revision Approval Cvcle No. Date Remarks 2 13 0 3/22/90 Original Unit 2 Core Operating Limits Report, distributed with Technical Specification Revision 92.

1 14 0 3/22/90 Original Unit 1 Core Operating Limits Report, distributed with Technical Specification Revision 92.

1 7/27/90 Incorporated expanded V(z) curves.

2 9/27/90 Clarified rod insertion limit curve applicability.

3 211 1/91 Incorporated revised FQof2.45 as a result of NRC approval of Westinghouse Topical Report WCAP-10924-P-A, Volume I , Addendum 4, October 1990.

0 - Not used.

1 9/27/90 Updated to Unit 2 Cycle 14, incorporated expanded V(z) curves and clarified rod insertion limit curve applicability.

2 211 1/91 Incorporated revised FQof2.45 as a result of NRC approval of Westinghouse Topical Report WCAP-10924-P-A, Volume 1, Addendum 4, October 1990.

2 15 0 3/9/92 Updated to Unit 2 Cycle 15 and clarified labeling of Figure 4. Clarified the actions to be taken if the nuclear enthalpy rise hot channel factor exceeds the Technical Specification limit.

1 16 0 12/28/92 Updated to Unit 1 Cycle 16, removed V(z) curves and replaced them with list of bounding V(z) values for three ranges of exposures. I 2 16 0 12/8/93 Updated to Unit 2 Cycle 16. Removed the multiple V(z) curves and replaced them with a single figure with bounding V(z) curves for four ranges of exposures. lncorporated additional discussion related to V(z) and K(z).

Page 1 of 6 I

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Unit Revision Approval Cycle No. Date Remarks 2 16 1 11/3/94 The table containing the bounding V(z) values and Figure 2 updated to incorporate revised bounding V(z) values for the exposure range of 14-21.5 GWDIMTU. Figures 3 through 6 re-formatted.

1 17 0 6/17/94 Updated to Unit 1 Cycle 17. Removed the list of bounding V(z) values and replaced it with multiple V(z) curves. lncorporated additional discussion related to V(z) and K(z).

2 17 0 6/2/95 Updated to Unit 2 Cycle 17. Incorporated Table 1 and expanded Figure 2 with updated bounding V(z) values.

1 18 0 2/7/96 Updated to Unit 1 Cycle 18, Incorporated revised FA, limit of 1.77. lncorporated Table 1 and updated Figure 2 with revised bounding V(z) values.

2 18 0 2127/97 Updated to Unit 2 Cycle 18. Revised FA, limit to I.77. Updated Table 1 and Figures 2a through 2e with revised bounding V(z) values.

lncorporated new Figures 2f and 2g with additional bounding V(z) values.

1 19 0 9/25/97 Updated to Unit 1 Cycle 19. Updated Table 1 and Figures 2a through 2f with revised bounding V(z) values.

2 19 0 12/17/98 Updated to Unit 2 Cycle 19. Updated Table 1 and Figures 2a through 2d with revised bounding V(z) values. Deleted Figures 2e, 2f and 2g.

1 20 0 5/13/99 Updated to Unit 1 Cycle 20. Updated Table 1 and Figures 2a through 2f with revised bounding V(z) values.

8/4/00 Technical Specification Amendment 151:

Relocate shutdown margin (SDM) requirements from Tech Specs and incorporate additional SDM requirements for Modes 3-6 from revised analysis of Uncontrolled Dilution event.

Page 2 of 6

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Unit Revision Approval Cycle No. Date Remarks Updated to Unit 2 Cycle 20. Updated Table 1 and Figures 2a through 2d with revised bounding V(z) values. Added new Table 2 and Figures 2e, 2f and 2g with additional bounding V(z) values.

Added references to Tables 1 and 2 and to Figures 2e, 2f and 2g to discussion of heat flux hot channel factor limits. Added discussion clarifying applicability of axial flux difference limits when using Tables 1 and 2 and Figures 2a through 2g. Added discussion of two tier V(z) curve presented in Table 2 and Figure 2g.

1 8/4/00 Technical Specification Amendment 142:

Relocate shutdown margin (SDM) requirements from Tech Specs and incorporate additional SDM requirements for Modes 3-6 from revised analysis of Uncontrolled Dilution event.

20 2 9/1/00 Revised to change axial flux difference target band.

21 0 1/31/01 Updated to support refueling activities associated with Unit 1 Cycle 21. Revision 0 of the Unit 1 Cycle 21 COLR had to be issued prior to confirming the applicability of the LOCA analysis.

Therefore, Revision 0 of the Unit 1 Cycle 21 COLR does not contain all of the operating limits necessary to support operation of Unit 1 Cycle 21.

21 1 2/19/01 Updated to Unit ICycle 21. Updated Tables 1 and 2 and Figures 2a through 2f with revised bounding V(z) values.

21 2 10/02/02 Revised to support License Amendment 158 changes, including revision of all references to TS, revision of FQsymbols, addition of Table 4, ITC limits, DNB limits and refueling boron concentrations.

21 0 2/06/02 Updated to Unit 2 Cycle 21.

Page 3 of 6

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Unit Revision Approval Cycle No. Date Remarks 2 21 1 10/02/02 Revised to support License Amendment 149 changes, including revision of all references to TS, revision of FQsymbols, addition of Table 4, ITC limits, DNB limits and refueling boron concentrations. Also revised to include an additional V(z) curve to give greater FQmargin between 13.0 and 16.0 GWdIMTU.

1 22 0 11/25/02 Updated to Unit ICycle 22. Updated Tables 1 and 2 and Figures 2a through 2f with revised bounding V(z) values. Incorporated new Figure 2g with additional bounding V(z) values.

Updated Table 3 with revised minimum shutdown margin limits. Deleted and revised text to eliminate duplication with the Technical Specifications and the Bases.

0 9/19/03 Updated to Unit 2 Cycle 22. Updated Tables 1 and 2. A reduced number of exposure ranges were calculated in Table I , therefore new Figures 2a through 2e with revised bounding V(z) values replaced Figures 2a through 2f. New Figure 2f replaced Figure 2g for the 2 tier band bounding V(z) values. Updated Table 3 with revised minimum shutdown margin limits. Deleted and revised text to eliminate duplication with the Technical Specifications and the Bases.

1 22 1 7/6/04 Revision to incorporate Westinghouse Safety Analysis Transition per LA 1621153. Revision 1 contains transitional values for the OPIOT AT Trip setpoints that will be used while the physical changes are implemented.

2 22 1 7/6/04 Revision to incorporate Westinghouse Safety Analysis transition per LA 1621153. Revision 1 contains transitional values for the OPIOT AT Trip setpoints that will be used while the physical changes are implemented.

2 22 2 7/12/04 Revised Fq limit from 2.4 to 2.5. Removed OP and OT delta-T setpoints based on NMC methodology and replaced with Westinghouse developed setpoints.

Page 4 of 6

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Unit Revision Approval Cycle No. Date Remarks 22 2 7116/04 Revised Fq limit form 2.4 to 2.5. Removed OP and OT delta-T setpoints based on NMC methodology and replaced with Westinghouse developed setpoints.

23 0 10/20/04 Updated to Unit 1 Cycle 23.

23 0 - Not used due to core redesign.

23 1 5/19/05 Updated to Unit 2 Cycle 23 and to support redesign of Unit 2 Cycle 23 core.

1 711 1/05 Revised ITC upper limit from < 0 pcmPF for power levels > 70% RTP to less than a line that slopes linearly from 0 pcmIoFat 70% RTP to -2.9 pcm/"F at 100% RTP. Revised the title of Figure 3 to reference T.S. 3.1.4 Condition B and revised the title of Figure 4 to reference T.S. 3.1.4 Condition A. Added references 24 and 25 to include the 50.59 screenings written to issue revision 1.

24 0 5/10/06 Updated to Unit 1 Cycle 24.

24 1 8/7/06 Updated Table 3 to reflect the correct FqW(z) penalty factors.

24 0 11/26/06 Updated to Unit 2 Cycle 24 Modes 5 and 6.

24 1 12/6/06 Updated to Unit 2 Cycle 24 for Modes 1-6.

24 2 9/4/07 Revised to support LA-1791169. Revised reference 24 to include the revision number (revision 0) and the correct date of the report (January 2005). Revised references 6a, 6b, 6c, and 8 to say 'Deleted.' These references referred to the old LBLOCA methodology and model.

24 2 211 1/08 Updated Table 1 to reflect correct Shutdown Margin Requirements and added Figures 6A through 6H.

Page 5 of 6

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Unit Revision Approval Cycle No. Date Remarks 2 24 3 211 1/08 Updated Table 1 to reflect correct Shutdown Margin Requirements and added Figures 6A through 6H.

1 25 0 2/24/08 Updated to Unit 1 Cycle 25 1 25 I 5/28/08 Updated Table 2 to reflect the correct W(z) at a burnup of 150 MWdIMTU and a core height of 6.20 feet 2 25 0 9/26/08 Updated for Unit 2 Cycle 25 1 26 0 9/24/09 Updated for Unit 1 Cycle 26 2 26 0 513110 Updated for Unit 2 Cycle 26 2 26 1 5117110 Updated to include part power W(z) factors 1 26 1 912110 Updated for second set of W(z) factors 2 26 2 9130110 Updated for Measurement Uncertainty Recapture power uprate to 1677 MWth and for a second set of W(z) factors 1 26 2 9130110 Updated for Measurement Uncertainty Recapture power uprate to 1677 MWth 1 26 3 12117110 Updated SDM in Table 1 for Mode 2 to say 1.9.

1 27 0 515111 Updated for Unit 1 Cycle 27 1 27 1 61211 1 Updated for Unit 1 Cycle 27 Modes 1 through 6 2 27 0 3/28/12 Updated for Unit 2 Cycle 27 Page 6 of 6

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 2 CYCLE 27 REVISION 0 Reviewed By: N&

Mark Brossart d Date: 3,/$h0I2 Supervisor, Reactor Systems Supervisor, PWR Analyses

, Approved By: Date:

Note: This report is not past of the Technical Specifications This report is referenced in the Technical Specifications Page 1 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 2 - CYCLE 27 REVISION 0 I This report provides the values of the limits for Unit 2 Cycle 27 as required by Technical Specification 5.6.5. These values have been established using NRC approved methodologies and are established such that all applicable limits of the safety analysis are met. The Technical Specifications affected by this report are listed below:

Reactor Core SLs Shutdown Margin (SDM)

Isothermal Temperature Coefficient (ITC)

Shutdown Bank Insertion Limits Control Bank Insertion Limits Physics Tests Exceptions - MODE 2 Heat Flux Hot Channel Factor (FQ(z))

Nuclear Enthalpy Rise Hot Channel Factor (F&)

Axial Flux Difference (AFD)

Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Technical Specification Table 3.3.1-1 (Note 1 and Note 2)

RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Boron Concentration

1. 2.1.1 Reactor Core Safety Limits The Reactor Core Safety Limits are shown in Figure 1.

Reference Technical Specification 2.1.1.

2. 3.1.1 Shutdown Margin Requirements The Minimum Shutdown Margin requirements are shown in Table 1 Reference Technical Specification 3.1.1.

Page 2 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0

3. 3.1.3 Isothermal Temperature Coefficient (ITC)

The ITC Upper limit is:

a. < 5 p c d ° F for power levels < 70% RTP; and

b. a line which slopes linearly from

i. 0 p c d ° F at a power level = 70% RTP to ii. -1.5 p c d ° F at a power level = 100% RTP The ITC Lower limit is:

a. -43.15 p c d ° F Reference Technical Specification 3.1.3.

4. 3.1.5 Shutdown Bank Insertion Limits The shutdown rods shall be fully withdrawn.

Reference Technical Specification 3.1-5.

5. 3.1.6 Control Bank Insertion Limits The control rod banlts shall be limited in physical insertion as shown in Figures 2,3, and 4.

The control rod banlts withdrawal sequence shall be Bank A, Bank B, Banlt C, and finally Bank D.

The control rod banks shall be withdrawn maintaining a 128 step tip-to-tip distance.

Reference Technical Specification 3.1.6.

6. 3.1.8 Physics Tests Exceptions - MODE 2 Minimum Shutdown Margin requirements during physics testing are shown in Table 1.

Reference Technical Specification 3.1.8.

Page 3 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0

7. 3.2.1 Heat Flux Hot Channel Factor ( F Q 0 The Heat Flux Hot Channel Factor shall be within the following limits:

CFQ = 2.50 K(Z) is a constant value = 1.0 at all elevations.

The HFP W(Z) values are provided in Table 2 and Table 4.

The W(Z) values in Table 2 are only applicable to Figure 5.

The W(Z) values in Table 4 are only applicable to Figure 6.

The data in Tables 2 and 4 should be used independently; cross interpolation or extrapolation between W(Z) sets is prohibited.

The Part Power W(Z) values for 75%<P<85% and Part Power W(Z) values for 85%<P<95% are provided in Table 6. The W(Z) values in Table 6 are only applicable to Figure 5.

The F ~ ~ ( Penalty z) Factors associated with Figure 5 and Table 2 are provided in Table 3.

The F ~ ~ ( Penalty z) Factors associated with Figure 6 and Table 4 are provided in Table 5.

The Axial Flux Difference (AFD) Band in Figure 6 is more restrictive that the AFD Band in Figure 5. Prior to switching from Figure 6 to Figure 5, F ~ ~ ( must z ) be confilmed to meet Technical Specification requirements by one of the following methods:

1. Confirm F ~ ~ ( meets z ) the Technical Specification Limit with the Table 2 W(Z) values for the most recent surveillance performed.

2. Perform a new surveillance and confirm F ~ ~ ( meets z ) the Technical Specification Limit with the Table 2 W(Z) values.

The HFP W(Z) values are generated assuming that they will be used for full power surveillance. When a past power surveillance is performed from BOC through 150 MWd/MTU and at a power level specified for Table 6, the W(Z) values provided in Table 6 should be used. When a part power surveillance is performed after 150 MWDIMTU, or at a power level other than those specified for Table 6, the HFP W(Z) values in Table 2 or Table 4 should be used.

Page 4 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 W(Z) values should be multiplied by the factor 1/P, when P > 0.5. When P is 5 0.5, the W(Z) values should be multiplied by the factor 1/(0.5), or 2.0. This is consistent with the adjustment in the FQ(Z)limit at part power conditions.

Reference 10 provides the basis for multiple sets of W(Z) curves.

Reference Technical Specification 3.2.1.

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor ( ~ ~ 3 The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limits:

FAHI 1.77 x [ l + 0.3(1- P)] for all 422V+ type fuel assemblies, and FA"I 1.66 x [1 + 0.3(1- P)] for all OFA type fuel assemblies where: P is the fraction of RATED THERMAL POWER at which the core is operating.

Reference Technical Specification 3.2.2.

3.2.3 Axial Flux Difference (AFD)

The indicated axial flux difference shall be maintained within the allowed operational space defined by Figure 5 or the more restrictive operational space as defined by Figure 6.

Both Figures 5 and 6 can be used any time during the cycle.

Prior to switching to the more restrictive AFD envelope (Figure 6), it should be confirmed that the plant is within the specified AFD envelope.

Reference Technical Specification 3.2.3.

Page 5 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0

10. 3.3.1 Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Technical Specification Table I 3.3.1-1 (Note 1 and Note 2);

Overtemperature AT Setpoint The Overtemperature AT setpoint parameter values are:

Indicated AT at RATED THERMAL POWER (%), OF Average temperature, OF 560.0 OF Pressurizer Pressure, psig 2235 psig 1.17 0.014 1°F 0.00100 /psi 30 seconds 4 seconds A function of the indicated difference between top and bottom detectors of the power range nuclear ion chambers. Selected gains are based on measured instrument response during plant startup tests, where q, and qb are the percent power in the top and bottom halves of the core respectively, and qt + qbis total core power in percent of RATED THERMAL POWER, such that (a) For q, - qb within -13, +8 % f(A1) = 0 (b) For each percent that the magnitude of q, - qb exceeds +8% the AT trip setpoint shall be automatically reduced by an equivalent of 1.73 % of RATED THERMAL POWER.

(c) For each percent that the magnitude of qt - qb exceeds -13 % the AT trip setpoint shall be automatically reduced by an equivalent of 3 346

% of RATED THERMAL POWER.

Overpower AT Setpoint The Overpower AT setpoint parameter values are:

AT0 = Indicated AT at RATED THERMAL POWER (%), OF T = Average temperature, O F T' = 560.0 OF K4 I 1.11 Ks = 0.0275/"F for increasing T; 0 for decreasing T K6

= 0.002I0Ffor T > T' ; 0 for T 5 T' 73 = 10 seconds Page 6 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0

11. 3.4.1 RCS Pressure. Temperature. and Flow - Departure from Nucleate Boiling ( D M ) Limits The DNB Limits are:

Pressurizer pressure limit = 2190 psia RCS average temperature limit = 564OF RCS total flow rate limit = 178,000 gpm Reference Technical Specification 3.4.1.

12. 3.9.1 Refueling Boron Concentration The boron concentration of the reactor coolant system and the refueling cavity shall be sufficient to ensure that the more restrictive of the following conditions is met:

a) Keff 20.95 b) 2000 ppm c) The Shutdown Margin specified in Table 1 Reference Technical Specification 3.9.1, Page 7 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 REFERENCES (NRC Approved Methodologies for COLR Parameters)

1. NSPNAD-8101-A, "Qualification of Reactor Physics Methods for Application to Prairie Island,"

Revision 2, October 2000.

2. NSPNAD-8 102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units," Revision 7, July 1999.

3. NSPNAD-97002-PA, "Northern States Power Company's "Steam Line Break Methodology,"

Revision 1, October 2000.

4. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July, 1985.

5.a WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," August, 1985.

5.b WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," Addendum 2 Revision 1, July 1997.

6. Not used.

7. WCAP-10924-P-A, Volume 1, Revision 1, and Volume 2, Revision 2, "Westinghouse Large Break LOCA Best Estimate Methodology," September 2005.

8. XN-NF-77-57-(A), XN-NF-77-57, Supplement 1 (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase 11," May 1981.

9. WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report: W-COBRAJTRAC2-Loop Upper Plenum Injection Model Update to Suppoi-t Z I R L O ~Cladding

~ Options," February 1994.

10. NSPNAD-93003-A, "Prairie Island Units 1 and 2 Transient Power Distribution Methodology,"

Revision 0, April 1993.

11. NAD-PI-003, "Prairie Island Nuclear Power Plant Required Shutdown Margin During Physics Tests," Revision 0, January 2001.

12. NAD-PI-004, "Prairie Island Nuclear Power Plant F ~ ~ ( Penalty z) With Increasing [F'~(z) / K(Z)

Trend," Revision 0, January 200 1.

13. WCAP- 10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification," February 1994.

zIRLo@ is a registered trademark of Westinghouse Electric Company LLC in the United States and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

I Page 8 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.

WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989.

WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report," January 1999.

WCAP-7588 Rev. I-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.

WCAP-7908-A, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a U 0 2 Fuel Rod," December 1989.

WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.

WCAP-7979-P-A, "TWINKLE - A Multidimensional Neutron Kinetics Computer Code,"

January 1975.

WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.

WCAP-11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.

WCAP-11596-P-A, "Qualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.

WCAP-12910 Rev. 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.

WCAP-14565-P-A, "VPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.

WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.

WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," Revision 0, January 2005.

Caldon, Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM dTMSystem," Revision 0, March 1997.

Caldon, Inc. Engineering Report-157P, "Supplement to Topical Report ER-8OP: Basis for a Power Uprate With the LEFM dTMCheck or CheckPlusTMSystem," Revision 5, October 200 1.

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.

Page 9 of 24

Core Operating Limits Report I Unit 2, Cycle 27 Revision 0 1 3 1. WCAP-12610-P-A & CENPD-404-P-A,Addendum 1-A, "Optimized ZIRLOTM,"July 2006.

1 32.

50.59 Evaluation 1094, Revision 0 "Unit 2 Cycle 27 Core Reload Modification."

Optinxized ZIRLOT" is a trademarlc of Westinghouse Electric Company LLC in the United States and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

Page 10 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Table 1 Minimum Required Shutdown Margin, %Ap i

L l U l l l U b l U L L l L U l S L l l S I U l l L f J D I\UIIIllIIF)

Mode 1*

0-1 Pump 2 Pumps 3 Pumps 0 - 21000 MWdMTU I Mode 2*

I 0-1 P u m ~ 2Pum~s 3Pum~s Physics Testing in Mode 2

1 0 - 21000 Mwd/MTu 0-1 Pump 0.5 2 Pumps 0.5 3 Pumps 0.5 1 Mode 3 T,,,, 2 520 OF (Most Reactive Rod Out)

I 0-1 Pump 2 Pumps 3 Pumps II 0 - 21000 MWdMTU 2.0 2.0 2.0 I Mode 3 350 OF I T,,,, < 520°F (Most Reactive Rod Out) 0-1 Pump 2 Pumps 3 Pumps 0 MWd/MTU 2.0 2.0 2.5 1 13000 MWdMTU 21000 M W d N T U 2.0 2.0 2.0 2.0 2.0 2.0 I Mode 4 200 OF < To,,,< 350°F (Most Reactive Rod Out) I 0-1 Pump 2 Pumps 3 Pumps 0 MWdMTU 2.0 4.0 6.0 13000 M W d N T U 2.0 2.5 4.0 21000 MWd/MTU 2.0 2.0 2.5 Operational Mode Definitions, as per TS Table 1.1-1.

• For Mode 1 and Mode 2 with Keff>1.0, the lninimuin shutdown inargin requirements are provided by the Rod Insertion Limits.

• • Charging pump(s) in service only pertains to steady state operations. It does not include transitory operations. For example, operations such as starting a second charging pump in order to secure the operating pump would fall under the one pump in service colunln Page 11 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Table 1, Continued Minimum Required Shutdown Margin, %Ap Number of Charging Pumps Running*"

Mode 5 68°F I T,,, I 200°F (Most Reactive Rod Out) 0-1 Pump 2 Pumps 3 Pumps 0 MWd/MTU*** 2.5 4.5 7.0 I Mode 6 68°F I T,,,, < 200°F (ARI)

I 0-1 Pump 2 Pumps 3 Pumps I OMWd/MTU*** 5.129 5.129 7.0 1 13000 MWdMTU 5.129 5.129 5.129 II Mode 6 68°F I T,,,, < 200°F (ARO)

I 0-1 Pump 2 Pumps 3 Pumps I OMWd/MTU*** 5.129 5.5 8.5 1 13000 MWd/MTU 5.129 5.129 6.5 21000 MWd/MTU 5.129 5.129 5.129 Operational Mode Definitions, as per TS Table 1.1-1.

• • Charging pump(s) in service only pertains to steady state operations. It does not include transitory operations. For example, operations such as starting a second charging pump in order to secure the operating pump would fall under the one pump in service colunm.

( *** These values are also applicable for the Unit 2 Cycle 26 end of cycle.

Page 12 of 24

Core Operating Limits Report Unit 2 , Cycle 27 Revision 0 I

• I Linear extrapolation based on a line between 16,000 MWDIMTU and 19,000 MWDIMTU is adequate for addressing burnup beyond 19,000 MWDIMTU.

Page 13 of 24

Core Operating Limits Report I Unit 2,Cycle 27 Revision 0 I

I Table 2 (cont.) - W(z) Values associated with Figure 5 (Top 10% and Bottom 8% excluded)*

I Height BU [MWdlMTU]

I [ftl 150 6000 9000 12000 14000 16000 19000 I A 0 = -2.63 A 0 = -3.56 A 0 = -4.73 A 0 = -5.41 A 0 = -2.49 A 0 = 0.41 A 0 = 0.66 I 38 7.40 1.1480 1.1792 1.1827 1 .I982 1.1845 1.1852 1.1995 I 39 7.60 1.1543 1.1846 1.I918 1 ,2075 1.1905 1.1850 1.1989 I 40 7.80 1.I610 1.1890 1 .I995 1.2153 1.1948 1 .I831 1 .I963 I

I I

I I

I I

I I

I I

I I

I I

I I 57 11.20 1.0000 1 .oooo 1 .oooo 1 .oooo 1 ,0000 1 .oooo 1 .oooo I 58 11.40 1.0000 1 .OOOO .

1 0000 1 .OOOO 1 .OOOO 1 .OOOO 1 .OOOO I 59 11.60 1.0000 1 .OOOO 1 .OOOO 1 .OOOO 1 .OOOO 1 .oooo .

1 0000 I 60 11.80 1.0000 1 .OOOO .

1 0000 1 . 0000 .

1 0000 1 .OOOO 1 .OOOO I * [TOP] 61 12.00 1 ,0000 1 .OOOO 1 .OOOO 1 .OOOO 1 .OOOO 1 ,0000 1 .OOOO Linear extrapolation based on a line between 16,000 MWDIMTU and 19,000 MWDIMTU is adequate for addressing burnups beyond 19,000 MWDIMTU.

Table 3 FWQ(2)Penalty Factor associated with Figure 5 and Table 2 I Cycle Burnup (MWDMTU) FWo(z)Penalty Factor I 0 1.0200 14465 1.0200 I 14606 1.0206 I 14746 1.0226 I 14886 1.0239 I 15027 1.0230 21000 1.0200 Linear interpolation is adequate for intermediate cycle bunlups.

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Core Operating Limits Report I Unit 2, Cycle 27 Revision 0 I 19 3.60 1.0792 1 .0686 1 .0576 1 .0525 1 ,0831 1.1165 1.1221 I 20 3.80 1.0811 1 ,0749 1 ,0638 1 ,0599 1 .0880 1.1174 1.1283 I 21 4.00 1 ,0845 1 ,0807 1 ,0706 1 .0669 1 ,0924 1.1208 1.1335 I 22 4.20 1 ,0879 1 .0859 1 .0763 1 .0734 1 ,0961 1.1253 1.1380 I 23 4.40 1 .0906 1 ,0907 1.0817 1 ,0795 1 .0993 1.1284 1.1414 I

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• Linear extrapolation based on a line between 16,000 MWDIMTU and 19,000 MWDIMTU is adequate for addressing burnups beyond 19,000 MWDIMTU.

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Core Operating Limits Report I Unit 2, Cycle 27 Revision 0 Table 4 (cont.) - W(z) Values associated with Figure 6 (Top 10% and Bottom 8%

excluded)*

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• Linear extrapolation based on a line between 16,000 MWDIMTU and 19,000 MWDIMTU is adequate for addressing bnrnups beyond 19,000 MWDIMTU.

Table 5 FWQ(z)Penalty Factor associated with Figure 6 and Table 4 Cycle Burnup (MWDJMTU) ~ ' " ~ ( zPenalty

) Factor 0 1.0200 21000 1.0200 Linear interpolation is adequate for intermediate cycle busnups.

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Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Table 6 - W(z) Values associated with Figure 5 for Part Power Surveillances*

(Top 10% and Bottom 8% excluded)

• W(z) values only valid for core average burnups 5 150 MWdIMTU.

80% of full power W(z) values are applicable for powers 75% 5 P < 85%.

90% of full power W(z) values are applicable for powers 85% 5 P < 95%.

$ Rod insertioll is given as a target value. Use control rods as necessary to control to target AO.

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Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Table 6 (cont.) - W(z) Values associated with Figure 5 for Part Power Surveillances*

(Top 10% and Bottom 8% excluded)

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• W(z) values only valid for core average bu~nups5 150 MWdIMTU.

80% of full power W(z) values are applicable for powers 75% 5 P < 85%.

90% of full power W(z) values are applicable for powers 85% 5 P < 95%.

$ Rod insertion is given as a target value. Use control rods as necessary to control to target AO.

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Core Operating Limits Repoi-t Unit 2, Cycle 27 Revision 0 Figure 1 Reactor Core Safety Limits 0 0.2 0.4 0.6 0.8 I 1.2 1.4 Fraction of 1677 MWt Rated Thermal Power Page 19 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Figure 2 Rod Insertion Limit, 128 Step Tip-to-Tip 0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:

BankD= (150 163) * (P- 100)+ 185 Bank C = (150 163) * (P - 100) + 185 + 128 NOTE: The top of the active fuel height con-esponds to 224 steps. The ARO parking position may be any position above 224 steps.

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Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Figure 3 Rod Insertion Limit, 128 Step Tip-to-Tip, One Bottomed Rod (Technical Specification 3.1.4, Condition B) 0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:

Bank D = (150 / 63) * (P - 90) + 224 Bank C = (150 / 63) * (P - 90) + 224 + 128 NOTE: The top of the active fuel height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.

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Core Operating Limits Repoi-t Unit 2, Cycle 27 Revisioll 0 Figure 4 Rod Insertion Limit, 128 Step Tip-to-Tip, One Inoperable Rod (Technical Specification 3.1.4, Condition A)

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0 20 40 60 80 100 Power Level, % of Rated Thermal Power Banli Positions Given By:

NOTE: The top of the active file1 height corresponds to 224 steps. The ARO parking positioil may be ally positioil above 224 steps.

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Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Figure 5 Flux Difference Operating Envelope associated with Table 2 and Table 6 Page 23 of 24

Core Operating Limits Report Unit 2, Cycle 27 Revision 0 Figure 6 Flux Difference Operating Envelope associated with Table 4 I

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