NLS2004143, Core Operating Limits Report, Cycle 22, Revision 2

Core Operating Limits Report, Cycle 22, Revision 2

ML043200649
Person / Time
Site:	Cooper
Issue date:	11/09/2004
From:	Fleming P Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2004143
Download: ML043200649 (24)

Text

Nebraska Public Power District Always there when you need us NLS2004143 November 9, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Core Operating Limits Report, Cycle 22, Revision 2 Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46

Reference:

I. Letter to U.S. Nuclear Regulatory Commission from Michael T. Coyle (Nebraska Public Power District) dated April 4, 2003, "Core Operating Limits Report" (NLS2003042)

2. Letter to U.S. Nuclear Regulatory Commission from Michael T. Coyle (Nebraska Public Power District) dated June 18, 2003, "Core Operating Limits Report, Cycle 22, Revision 1" (NLS2003061)

The purpose of this letter is to provide the Nuclear Regulatory Commission (NRC) the revised Core Operating Limits Report (COLR) for Cooper Nuclear Station (CNS) for Cycle 22. CNS Technical Specification 5.6.5.d requires that the COLR, including any midcycle revisions or supplements, be provided to the NRC upon issuance for each reload cycle. The Cycle 22 COLR, Revision 0 (Reference 1) contained an administrative error that was corrected with Revision I (Reference 2). The Cycle 22 COLR, Revision I (Reference 2) contained a technical error in that the Cycle 22 Rod Withdrawal Error (RWE) analysis was not bounded by the generic RWE analysis. The Cycle 22 COLR contained the incorrect statement that "the trip level settings associated with this MCPR [Minimum Critical Power Ratio] limit have been generically calculated and verified to bound the Rod Withdrawal Error Analysis for Cycle 22 operation."

This statement has been corrected in Revision 2. In accordance with 10 CFR 50.4(b)(1), we are also transmitting a copy of this COLR to the Regional Office and to the NRC Senior Resident Inspector.

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

NLS2004 143 Page 2 of 2 Should you have any questions regarding this matter, please contact Mr. Paul Fleming at (402) 825-2774.

Sincerely, 4r Paul V. Fleming Licensing Manager

/cb Enclosure cc: Regional Administrator w/enclosure USNRC Region IV Senior Project Manager w/enclosure USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/enclosure USNRC NPG Distribution w/o enclosure Records wv/enclosure

NLS2004 143 Enclosure ENCLOSURE CORE OPERATING LIMITS REPORT CYCLE 22, REVISION 2 COOPER NUCLEAR STATION DOCKET No. 50-298, DPR-46

COOPER NUCLEAR STATION CORE OPERATING LIMITS REPORT Cycle 22 Revision 2 Cycle 22, Revision 2 Signature Page Revision 2 I Preparer:

Print b QL4 \-wcA Sign/Date Reviewer:

Print Sign/Date RE Manager: /. /eu IS

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V - Signruvate Cycle 22, Revision 2 I

1.0 INTRODUCTION

The Core Operating Limits Report provides the limits for operation of the Cooper Nuclear Station for Cycle 22. It includes the limits for the Rod Block Monitor Upscale Set Point, Average Planar Linear Heat Generation Rate (APLHGR), and Minimum Critical Power Ratio (MCPR). In addition, this COLR also contains:

- MCPR limits for an inoperable Main Turbine Bypass System (one bypass valve inoperable)

- Power to flow map defining the Stability Exclusion Region

- Turbine Bypass System response time

- Maximum allowable LHGR If any of these limits is exceeded, action will be taken as defined in the Technical Specifications.

The core operating limit values have been determined using the NRC-approved methodologies given in References 1, 2, 10, and 11 and have been established such that all applicable plant safety analysis limits are met.

2.0 CORE OPERATING LIMITS Cooper Nuclear Station shall operate within the bounds of the below limits/values. The applicable Technical Specifications are referenced in each subsection.

2.1 Rod Block Monitor Upscale Set Point The Technical Specifications reflect a reference to Allowable Values for the Rod Block Monitor (RBM) upscale (power referenced) trip level setting, found in Reference 9, are as follows:

Lowest Rated Low Trip Set Intermed Trip Set High Trip Set MCPR Limit Point (LTSP) Point (ITSP) Point (HTSP)

(LPSP*P*IPSP) (IPSP<P*HPSP) (HPSP<P) 21.33 *1 17.0/125 5112.5/125 5107.5/125 LPSP, IPSP, and HPSP are the Low Power Set Point, Intermediate Power Set Point, and High Power Set Point, respectively.

The trip level settings associated with this MCPR limit have been verified to bound the Rod Withdrawal Error Analysis for Cycle 22 operation as found in Reference 5.

Technical Specification

Reference:

3.3.2.1 Cycle 22, Revision 2 2.2 Average Planar Linear Heat Generation Limits The most limiting lattice APLHGR value (excluding natural uranium) for each fuel bundle as a function of Planar Average Exposure, core power, and core flow is calculated by multiplying the value from Figures 1, 2, 3, 4, 5, and 6 by the smaller of the MAPLHGR Flow Factor, MAPFACF (Figure 7) or the Power-Dependent MAPLHGR Factor, MAPFACp, (Figure 8). APLHGR values determined with the SAFER/GESTR-LOCA methodology are given in References 2, 3, and 5 while MAPFACF and MAPFACp were determined in Reference 8.

The calculated maximum APLHGR (MAPLHGR) limits in Figures 1, 2,3,4, 5 and 6 are conservative values bounding all fuel lattice types (excluding natural uranium) in a given fuel bundle design. MAPLHGR limits for each individual fuel lattice design in a bundle design, as a function of axial location and average planar exposure, are determined based on the approved methodology referenced in Technical Specification 5.6.5 and loaded in the process computer for use in core monitoring calculations. The MAPLHGR values for these lattices, along with the axial location of each lattice in the bundle, are considered proprietary information by General Electric and are given in Reference 3 as a function of planar average exposure.

The MAPLHGR limits referred to above are for two recirculation loop operations.

For single loop operation, the limiting APLHGR value is multiplied by 0.77 for GE8x8 NB fuel (as can be found in Reference 5) and by 0.91 for GE14 fuel (as can be found in Reference 5).

Technical Specification

Reference:

3.2.1 and 3.4.1 2.3 Linear Heat Generation Rate Limit The limiting power density and maximum allowable Linear Heat Generation Rate (LHGR) referred to in Technical Requirements Manual Section T 3.2.1 is the design LHGR. The design LHGR for fuel type GE 8x8 NB is 14.4 kW/ft as found in Reference 12. The design LHGR for fuel type GE14 is 13.4 kWift as found in Reference 13.

2.4 Minimum Critical Power Ratio Limits The operating limit MCPR (OLMCPR) values are a function of core thermal power, core flow, fuel bundle, scram time (T), and fuel exposure. The scram time (T) is determined from CNS Procedure 10.9, Control Rod Scram Time Evaluation.

The OLMCPR values are as follows:

For core thermal power 2 25 percent and <30 percent of rated power, the OLMCPR is the power dependent MCPR (MCPRp) from Figure 9.

Cycle 22, Revision 2 For core thermal power 2 30 percent of rated power, the OLMCPR is the greater of either:

The applicable flow dependent MCPR (MCPRF) determined from Figure 10, or the appropriate scram time (T) dependent MCPR at rated power from Figures I 1, and 12, multiplied by the applicable power dependent MCPR multiplier (Kp) from Figure 9.

The appropriate scram time (T) dependent MCPR at rated power with One Turbine Bypass Valve Unavailable is shown in Figure 13.

The system response time for the Turbine Bypass System to be at 80% of rated bypass flow is 0.3 seconds.

For single recirculation loop operation, the OLMCPR is 0.02 greater than the two recirculation loop operation OLMCPR.

Technical Specification

References:

3.2.2, 3.4.1 and 3.7.7 2.5 Power/Flow Map The power/flow map defining the Stability Exclusion Region can be found as Figure 14. References 5 and 6 reflect the documents describing the current Cooper Nuclear Station power/flow map. The Stability Exclusion Region boundary is given by the equation IIW-iJB, W,-JJB 1

( I2[ WA-WJ WA-WE )2]

P =PBW where, P = a core thermal power value on the region boundary (% of rated),

W = the core flow rate corresponding to power, P, on the region boundary (% of rated),

PA = core thermal power at point A (% of rated),

PB = core thermal power at point B (%of rated),

WA = core flow rate at point A (% of rated), and WB = core flow rate at point B (% of rated).

Technical Specification

Reference:

3.4.1 Cycle 22, Revision 2

3.0 REFERENCES

1. NEDE-2401 I-P-A-14-US, June 2000, GeneralElectricStandardApplicationfor ReactorFuel.

2. NEDC-32687P, Revision 1, March 1997, CooperNuclear Station SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis.

3. Lattice Dependent MAPLHGR Reportfor Cooper Nuclear Station Reload 21, Cycle 22, 0000-0002-9865-MAPL, Revision 0.

4. Letter (with attachment), R.H. Buckholz (GE) to P.S. Check (NRC) dated September 5, 1980, Response to NRC Request for Information on ODYN Computer Model.

5. Supplemental Reload Licensing Reportfor CooperNuclear Station Reload 21, Cycle 22, 0000-0002-9865-SRLR, Revision 0.

6. GE-NE-A 13-00395-01, Class 1,November, 1996, Application of the "RegionalExclusion with Flow-BiasedAPRMNeutron Flux Scram " Stability Solution (Option 1-D) to the CooperNuclear Station, Licensing Topical Report.

7. Letter from James R. Hall (NRC) to G. R. Horn (NPPD) dated September 23, 1997, Approval of SAFERIGESTAR LOCA Analysisfor CooperNuclear Station (TACNO. M98293)

8. GE-NE-L12-00867-12, CooperNuclearStation MG Project Task 900: Transient Analysis, Revision 1, May 2000.

9. NEDC 98-024, Revision 3, May 2000, APRM- RBMSetpoint Calculation.

10. NEDO-31960-A and NEDO-31960-A Supplement 1,BWR Owner's Group Long-Term Stability Solutions Licensing Methodology.

(The approved revision at the time the reload analysis is performed.)

11. NEDE-23785-1 -P-A, The GESTR-LOCA andSAFER Modelsfor the Evaluationof the Loss-of-CoolantAccident, Volume 111, Revision 1, October 1984.

12. Nuclear Design Reportfor CooperNuclear Station Reload, 18, J11-03354-03, July 1998.

13. GE-NE-L12-00867-09-02, CooperNuclear Station MIG Project Task 407: SAFER/GESTR-LOCA Analysis, May 2000.

14. Letter from S. Shelton (GNF) to J.L. Lewis (NPPD) dated December 21, 2000, GE9B LHGR Relaxationfor CooperNuclear Station.

15. NEDE-31152P, GE FuelBundle Designs, December 1988 (As Revised)

16. NEDC-32538P-A, DeterminationofLimiting Cold Water Event

17. GE-NE-J] 103910-09-02P, CooperNuclearStation ECCS-LOCA Evaluationfor Cycle 21, August 2001.

18. GE-NE-J1 103910-09-01, CooperNuclear Station ECCS-LOCA Evaluationfor GEM4, August 2001.

19. NEDC-32687P, CooperNuclear Station SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, Revison 1, March 1997.

Cycle 22, Revision 2 CORE OPERATING LIMITS REPORT Figure I Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure with LPCI Modification and Bypass Holes Plugged, Bundle 2610, 3.93 w/o with 17GZ GEI4C Fuel 15.00 12.00 A nn% d

.uu Ift 6.00

-J a-

'C 3.00 0.00 4-0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 Planar Average Exposure (GWdlST)

DATA COORDINATES {Reference 3)

Planar Average Exposure MAPLHGR (GWDIST) (kW/ft) 0.00 9.34 0.20 9.43 1.00 9.58 2.00 9.79 3.00 10.00 4.00 10.23 5.00 10.45 6.00 10.61 7.00 10.74 8.00 10.88 9.00 11.01 10.00 11.13 11.00 11.23 12.00 11.32 13.00 11.39 14.00 11.41 15.00 11.42 17.00 11.24 20.00 10.92 25.00 10.39 30.00 9.86 35.00 9.37 40.00 8.87 45.00 8.35 50.00 7.82 55.00 5.86 57.05 4.84 57.08 4.83 57.41 4.88 57.42 4.87 Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure with LPCI Modification and Bypass Holes Plugged, Bundle 2568, 3.98 w/o with 16GZ GE14C Fuel 15.00 12.00 F

I 9.00 --=-=-

0~ 6.00 3.00 nnn 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Planar Average Exposure (GWdIST)

DATA COORDINATES (Reference 3)

Planar Averaae Exposure MAPLHGR (GWDIST) (kW/ft) 0.00 9.63 0.20 9.64 1.00 9.71 2.00 9.86 3.00 10.03 4.00 10.23 5.00 10.45 6.00 10.66 7.00 10.87 8.00 11.09 9.00 11.20 10.00 11.28 11.00 11.36 12.00 11.42 13.00 11.42 14.00 11.42 15.00 11.42 17.00 11.33 20.00 11.00 25.00 10.43 30.00 9.89 35.00 9.39 40.00 8.90 45.00 8.38 50.00 7.85 55.00 6.07 57.47 4.83 57.49 4.83 58.31 4.89 58.32 4.88 Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure with LPCI Modification and Bypass Holes Plugged, Bundle 2472, 3.79 wlo with 17GZ GE14C Fuel 15.00 1 12.00

• 9.00I I 6.00

-J 0.

<: 3.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Planar Average Exposure (GWdIST)

DATA COORDINATES (Reference 3)

Planar Average Exposure MAPLHGR (GWD/ST) (kW/ft) 0.00 9.25 0.20 9.32 1.00 9.45 2.00 9.62 3.00 9.80 4.00 9.99 5.00 10.18 6.00 10.38 7.00 10.52 8.00 10.62 9.00 10.74 10.00 10.86 11.00 10.96 12.00 11.04 13.00 11.10 14.00 11.13 15.00 11.04 17.00 10.85 20.00 10.56 25.00 10.10 30.00 9.67 35.00 9.25 40.00 8.81 45.00 8.31 50.00 7.75 55.00 5.52 56.21 4.91 56.48 4.93 56.83 4.83 56.90 4.83 Cycle 22, Revision 2 I

.9-

CORE OPERATING LIMITS REPORT Figure 4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure with LPCI Modification and Bypass Holes Plugged, Bundle 2380, 3.85 wlo with 14GZ GE14B Fuel

_ 15.00 1 12.00

£ 9.00 V I 6.00

-I

'S 3.00 I:

E 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Planar Average Exposure (GWdIST)

DATA COORDINATES (Reference 3)

Planar Average ExDosure MAPLHGR (GWDlST) (kW/ft) 0.00 9.02 0.20 9.11 1.00 9.26 2.00 9.49 3.00 9.71 4.00 9.92 5.00 10.05 6.00 10.16 7.00 10.24 8.00 10.31 9.00 10.38 10.00 10.46 11.00 10.54 12.00 10.63 13.00 10.74 14.00 10.83 15.00 10.80 17.00 10.69 20.00 10.46 25.00 10.05 30.00 9.64 35.00 9.19 40.00 8.69 45.00 8.14 50.00 7.51 55.00 6.85 56.86 6.59 56.91 6.59 57.87 6.63 57.93 6.65 Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure with LPCI Modification and Bypass Holes Plugged, Bundle 2299, 3.50 w/o with IOGZ1 GE8X8NB Fuel

_ 15.00 12.00 01VW.W-W - - -

- 9.00 I 6.00

-j a- 3.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Planar Average Exposure (GWd/ST)

DATA COORDINATES (Reference 3)

Planar Average Exposure MAPLHGR (GWDIST) (kWlft) 0.00 11.59 0.20 11.63 1.00 11.71 2.00 11.84 3.00 11.99 4.00 12.14 5.00 12.26 6.00 12.39 7.00 12.53 8.00 12.66 9.00 12.81 10.00 12.85 12.50 12.79 15.00 12.51 20.00 11.78 25.00 11.08 35.00 10.53 45.00 9.51 53.92 5.74 54.00 5.75 Cycle 22, Revision 2 I

- II -

CORE OPERATING LIMITS REPORT Figure 6 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus Exposure with LPCI Modification and Bypass Holes Plugged, Bundle 2205, 3.50 wlo with 10GZ GE8X8NB Fuel 15.00 12.00 _________-a--

CD 9.00

0. 6.00 IV 3.00 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Planar Average Exposure (GWdIST)

DATA COORDINATES (Reference 3)

Planar Average Exmosure MAPLHGR (GWD/ST) (kW/ft) 0.00 11.59 0.20 11.63 1.00 11.71 2.00 11.85 3.00 12.00 4.00 12.13 5.00 12.26 6.00 12.38 7.00 12.52 8.00 12.65 9.00 12.80 10.00 12.84 12.50 12.81 15.00 12.52 20.00 11.78 25.00 11.09 35.00 10.53 45.00 9.54 54.00 5.75 54.02 5.74 Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 7 Reference 8 Figure 3-11 Cooper Nuclear Station Flow-Dependent MAPLHGR Factor, MAPFAC(f) 1.1 1

0.9 107.5%

112.)

0.8 117.)

MAPLHGR(I)= MAPFAC(t) MAPLHGRstd U .MAPLHGRstd = Standard MAPLHGR Limits L 0.7 For Two Loop Operation MAPFAC() - The Minimum of either 1.0or{ AF-(WCt1OO)+ BF)

Wc = % Rated Core Flow 0.6 AF and B F are fuel dependent constants given below for GE Fuels through GE14:

Maximum Core Flow 0.5 (Y se)AF -BF 102.5 0.6784 0.4861 107.0 0.6758 0.4574 112.0 0.6807 0.4214 117.0 0.6886 0.3828 0.4 0.3 30 40 50 60 70 80 90 100 110 Core Flow (% Rated)

Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 8 Reference 8 Figure 3-9 Cooper Nuclear Station Power-Dependent MAPLHGR Factor MAPFAC(p) 1.00 0.90 0.80

____ ____ ____MAPHGRp)= MAPRAC(p) *MAR.HGRstd 0.70 MAPLHGRstd =Standard MAPU-HGR Lirrits For P<25%: No Thermal Lwrfts Moinitoring Requird 0.60 For 25% < P <30%

MAPFAC(p)=-0.53 + 0.005(P.30%)for Flow < 50%

_______MAPFACqp)=0.49 + 0.005(P.30%)for Flow > 50%

0.50 i-For 30% < P MAPFAC~p) 1.0 + 0.005224(P-100%)

0.40 _ /7 0.30 0.20 0 10 20 30 40 50 60 70 80 90 100 POWER (% Rated)

Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 9 Reference 5 Cooper Nuclear Station Power-Dependent MCPR Limits, Kp and MCPRp 4.0 Operating Limit MCPR (P) = Kp Operating Limit MCPR (100)

For P< 25%: No Thermal Umits Monitoring Required 3.5 No limits specified For 25% S P < P(Bypass): (P(Bypass) = 30%)

K(P) = [ K(Byp) + 0.0500(30% - P)j I OLMCPR(1 00)

K(Byp) = 2.24 for Flow < 50%

K(P) = (K(Byp) + 0.0519(30% - P)I I OLMCPR(100) 3.0 K(Byp) = 2.66 for Flow > 50%

ow > 50%k For 30% < P < 45%: K(P) = 1.28 + 0.0134 (45% - P)

For 45% < P < 60%: K(P)= 1.15 + 0.00867 (60% - P)

For 60% S P: K(P) = 1.0 + 0.00375 (100% - P) 2.5

- . - - - - .a . - 2,0)

Flow <- 50%

2.0 1.5 1.0 20 30 40 50 60 70 80 90 100 110 Power (% rated)

Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 10 Reference 5 Cooper Nuclear Station Flow-Dependent MCPR Limits, MCPR(f) 1.8 For W(C) (% Rated Core Flow) t40%

GEXL05 orrectio MCPR(F) a MAX(t.22.A(F) *W(C)1100 + B(F))

Max Flow- 117.0 A(F).-0.644 B(F) -1.843 1.7

\7'- Max Flow 112.0 A(F)--0.613 B(F) . 1.780 112.0% Max Flow- 107.0 A(F) * -0.597 B(F)

• 1.729 107.0% Max Flow = 102.5 A(F) -0.582 B(F) - 1.686 1.6 ------ GE14 40% Flow 102.5% For W(C) (% Rated Core Flow) < 40% For GEXL05 only MCPR(F) a (A(F)

• W(C) 1100 4 B(F)) I (1 + 0.0032 * (40 - W(C)))

1.5 IL C,

0.

All Ot erGEXL \

1.4

"'11\ I S"".

1.3 I 1 I ._

1.1 Il--

20 30 40 50 60 70 80 90 100 110 120 Core Flow (% Rated)

Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure I I Minimum Critical Power Ratio (MCPR) versus Tau (based on tested measured scram time as defined in Reference 4), All Fuel 1.47 -_

1.46 -_

1.45 -_

1.44 -_

1.43 -_

It. 1.42- -

t 1.41 -_

1.4 -_

1.39 -

1.38 -

1.37 -

1.36 -

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Tau Exposure Range: BOC22 to EOC22-2315 MWd/MT (2100 MWdlST) ICF Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 12 Minimum Critical Power Ratio (MCPR) versus Tau (based on tested measured scram time as defined in Reference 4), All Fuel 1.59 1.58 1.57 1.56 1.55 1.54 1.53 0: 1.52

a. 1.51 o 1.5

• 1.49 1.48 1.47 1.46 1.45 1.44 1.43 1.42 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Tau Exposure Range: EOC22-2315 MWdIMT (2100 MWd/ST) to EOC22 ICF Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 13 Minimum Critical Power Ratio (MCPR) versus Tau (based on tested measured scram time as defined in Reference 4), All Fuel 1.59 1.58 1.57 1.56 - -

1.55 1.54 1.53 1.52 0a 1.51 /--

0 1.5 1.49 1.48 1.47 1.46 . I 1.45 1.44 1.43 . . , *. . .I 1.42

).0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Tau Exposure Range: BOC22 to EOC22 ICFITBPOOS Cycle 22, Revision 2 I CORE OPERATING LIMITS REPORT Figure 14 Cycle 22 Stability Exclusion Region 110.000 100.000 -

90.000 -

80.000 -

Point A y 70.000 -

60.000-50.000 /

40.000 -

Point B 30.000-20.000 -

20 30 40 50 60 70 80 Core Flow (%)

Exclusion Region Endpoints Power (% rated) Flow (% rated)

A 77.1 49.7 B 36.8 32.5 Cycle 22, Revision 2 I I ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS©)

Correspondence Number: NLS2004143 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing & Regulatory Affairs Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None I PROCEDURE 0.42 l REVISION 15 l PAGE 18 OF 24