ML090970247

Submittal of Cycle 19 Core Operating Limits Report

ML090970247
Person / Time
Site:	Brunswick
Issue date:	04/01/2009
From:	Mentel P N Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 09-0034
Download: ML090970247 (57)

Text

Progress.

Energy APR 0 12009 SERIAL: BSEP 09-0034 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit No. 2 Docket No. 50-324/License No. DPR-62 Unit 2 Cycle 19 Core Operating Limits Report Ladies and Gentlemen:

Enclosed is a copy of the Core Operating Limits Report (COLR):for Brunswick Steam Electric Plant (BSEP), Unit 2, Cycle, 19 operation.

Carolina Power &-Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc. is providing the enclosed COLR in accordance with BSEP Unit 2 Technical Specification'5.6.5.d.

The COLR provided in Enclosure 1 supersedes the report previously submitted by letter dated July 17, 2008 (i.e., ADAMS Accession Number ML082190542).

By letter dated January 24, 2008 (i.e., ADAMSAccession Number ML0803.10843), CP&L provided a response to NRC questions concerning the license amendment request supporting the use of AREVA fuel. As part of the response to Question 1, CP&L stated that with the transmittal of the COLR for the firstUnit 2 outage that uses AREVAfuel, an information-only transmittal of the thermal-hydraulic designfuel cycle design, and, reload safety analysis reports would be provided.In Enclosure 2, CP&L is providing a copy of AREVA Report ANP-2729(P), Revision 0, Brunswick Unit 2 Thermal-Hydraulic Design-Report for ATRIUMr m-] o0Fuel Assemblies, dated July 2008. ANP-2729(P) contains information that AREVA considers proprietary, as defined in 10 CFR 2.390. AREVA, as the owner of the proprietary information, has executed the affidavit provided in Enclosure 3, which states that the proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure.

AREVA requests that the proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390. A non-proprietary version of ANP-2729(P), is provided in Enclosure 4.In Enclosure 5, CP&L is providinga copy of AREVA:Report'ANP-2727(P), Revision 0, Brunswick Unit 2 Cycle 19 Fuel Cycle Design, dated June 2008. ,ANP-2727(P) contains information that AREVA considers proprietary,as' defined, in 10 CFR 2.390. ,AREVA, as the owner of the proprietary information, has executed-the affidavit provided in Progress Energy Carolinas, Inc.Brunswick Nuclear Plant PO Box 10429 Southport, NC 28461 Document Control Desk BSEP 09-0034 / Page 2 Enclosure 6, which states that the proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure.

AREVA requests that the proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.3 90. A non-proprietary version of ANP-2727(P), is provided in Enclosure 7.In Enclosure 8, CP&L is providing a copy of AREVA Report ANP-2771(P), Revision 0, Brunswick Unit 2 Cycle 19 Reload Safety Analysis, dated January 2009. ANP-2771 (P)contains information that AREVA considers proprietary, as defined in 10 CFR 2.390.AREVA, as the owner of the proprietary information, has executed the affidavit provided in Enclosure 9, which states that the proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure.

AREVA requests that the proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390. A non-proprietary version ofANP-2771 (P), is provided in Enclosure 10.No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Gene Atkinson, Supervisor

-Licensing/Regulatory Programs, at (910) 457-2056.Sincerely, Phyllis N. Mentel Manager -Support Services Brunswick Steam Electric Plant Document Control Desk BSEP 09-0034 / Page 3 WRM/wrm

Enclosures:

1. Brunswick Unit 2, Cycle 19 Core Operating Limits Report, March 2009 2. AREVA Report ANP-2729(P), Revision 0, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUMT"M IO Fuel Assemblies, dated July 2008 (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)3. AREVA Affidavit Regarding Withholding ANP-2729(P), Revision 0, from Public Disclosure

4. AREVA Report ANP-2729(NP), Revision 0, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUMTrMLIO Fuel Assemblies, dated July 2008 5. AREVA Report ANP-2727(P), Revision 0, Brunswick Unit 2 Cycle 19 Fuel Cycle Design, dated June 2008 (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)6. AREVA Affidavit Regarding Withholding ANP-2727(P), Revision 0, from Public Disclosure

7. AREVA Report AREVA Report ANP-2727(NP), Revision 0, Brunswick Unit 2 Cycle 19 Fuel Cycle Design, dated June 2008 8. ANP-2771 (P), Revision 0, Brunswick Unit 2 Cycle] 9 Reload Safety Analysis, dated January 2009 (Proprietary Information

-Withhold from Public Disclosure in Accordance With 10 CFR 2.390)9. AREVA Affidavit Regarding Withholding ANP-2771 (P), Revision 0, from Public Disclosure

10. ANP-2771 (NP), Revision 0, Brunswick Unit 2 Cycle l9 Reload Safety Analysis, dated January 2009 Document Control Desk BSEP 09-0034 / Page 4 cc (with Enclosures 1 through 10): U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A)11555 Rockville Pike Rockville, MD 20852-2738 cc (without enclosures):

Chair -North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 BSEP 09-0034 Enclosure 1 Brunswick Unit 2, Cycle 19 Core Operating Limits Report, March 2009 r Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 1, Revision 0 BRUNSWICK UNIT 2, CYCLE 19 CORE OPERATING LIMITS REPORT March 2009 Prepared By: Verified By: Approved By: Pribyl, David 2009.03.27 13:56:03 -04'00'David Pribyl BWR Fuel Engineering

-Senior Engineer Westmoreland, Gregory 2009.03.27 14:21:49 -04'00'Greg Westmoreland BWR Fuel Engineering

-Lead Engineer Blom, Michael Supervisor Approval: 2009.03.27 15:28:14 -04'00'Michael Blom BWR Fuel Engineering

-Supervisor Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 2, Revision 0 LIST OF EFFECTIVE PAGES Pa-ge(s)1-48 Revision 0 This document consists of 48 total pages.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design CaIc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 3, Revision 0 TABLE OF CONTENTS Sublect Pae C o v e r .................................................................................................

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1 List of Effective Pages ..........................................................................................................................

2 T a b le o f C o nte nts ...............................................................................................................................

3 L is t o f T a b le s .......................................................................................................................................

4 L is t o f F ig u re s .....................................................................................................................................

5 Nomenclature

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6 Introduction and Summary .....................................................................

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8 A P L H G R L im its ......................................................................

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9 MCPR Limits .....................................................................................................

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9 L H G R L im its ...............................................................................................................................

...... 1 0 P B D A S e tp o ints ................................................................................................................................

1 1 R B M S e tp o ints .........................................................

7*... .... .... .... .....................................................

12 Equipment Out-of-Service

........................................................... , ......................................................

12 Single Loop Operation

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12 Inoperable Main Turbine Bypass System ..........................................................................................

13 Feedwater Temperature Reduction

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13 R e fe re n c e s ..........................................................................................................................................

14 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design CaIc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 4, Revision 0 CAUTION References to COLR Figures or Tables should be made using titles only; Figure and Table numbers may change from cycle to cycle.LIST OF TABLES Table Title Paqe Table 1: R BM System Setpoints

..............................................................................................

16 Table 2: RBM O perability Requirem ents .................................................................................

17 T able 3: P B D A S etpoints ......................................................................................................

..18 Table 4: Exposure Basis for Brunswick Unit 2 Cycle 19 Transient Analysis ...............

19 Table 5: Power-Dependent M CPRp Lim its ................................................................................

20 NSS Insertion Times -BOC to < NEOC Table 6: Power-Dependent M CPRp Lim its ................................................................................

21 TSSS Insertion Times -BOC to < NEOC Table 7: Power-Dependent M CPRP Lim its ................................................................................

22 NSS Insertion Times -BOC to < EOCLB Table 8: Power-Dependent M CPRp Lim its ................................................................................

23 TSSS Insertion Times -BOC to < EOCLB Table 9: Power-Dependent M CPRP Lim its ................................................................................

24 NSS Insertion Times -BOC to < MCE (FFTR/Coastdown)

Table 10: Power-Dependent M CPRp Lim its ................................................................................

25 TSSS Insertion Times -BOC to < MCE (FFTR/Coastdown)

Table 11: Flow-Dependent M CPRf Lim its ..................................................................................

26 Table 12: ATRIUM-10 Steady State LHGRss Limits ..................................................................

27 Table 13: GE14 Steady-State LHGRss Limits ...................................

28 Table 14: ATRIUM-10 Power-Dependent LHGRFACp Multipliers

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29 NSS Insertion Times -BOC to < EOCLB Table 15: ATRIUM-10 Power-Dependent LHGRFACp Multipliers

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30 TSSS Insertion Times -BOC to < EOCLB Table 16: ATRIUM-10 Power-Dependent LHGRFACp Multipliers

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31 NSS Insertion Times -BOC to < MCE (FFTR/Coastdown)

Table 17: ATRIUM-10 Power-Dependent LHGRFACp Multipliers

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32 TSSS Insertion Times -BOC to < MCE (FFTR/Coastdown)

Table 18: ATRIUM-10 Flow-Dependent LHGRFACf Multipliers

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33 Table 19: ATRIUM-10 Steady-State MAPLHGRss Limits ..........................................................

34 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Caic. No. 2B21-1293 B2C19 Core Operating Limits Report Page 5, Revision 0 Table 20: GE14 Steady-State MAPLHGRss Limits ..............................................................

....... 35 GE14-P1ODNAB425-3G7.0/14G6.0/l G2.0-10OT-150-T-2574 (only)Table 21: GE14 Steady-State MAPLHGRss Limits ....................................

36 G E14-P1ODNAB439-12G6.0-10OT-150-T-2575 (only)Table 22: G El4 Steady-State MAPLHG Rss Lim its ..........................................................................

37 GE14-PlODNAB413-16GZ-10OT-150-T-2660 (only)Table 23: [Not Used] .........................................................

38 Table 24: GE14 Steady-State MAPLHGRss Limits .....................................................................

39 GE14-P1ODNAB407-16GZ-10OT-150-T-2853 (only)Table 25: GE14 Steady-State MAPLHGRss Limits ...................................

40 GE 14-PIODNAB425-18GZ-10OT-150-T-2854 (only)Table 26: GE14 Power-Dependent MAPFACp Multipliers

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41 Table 27: GE14 Flow-Dependent MAPFACf Multipliers...................................

42 LIST OF FIGURES Figure Title or Description Page Figure 1: Stability O ption III Power/Flow M ap ...........................................................................

43 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability O ption III Power/Flow M ap ....................

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44 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability O ption III Power/Flow M ap ............................................................................

45 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability Option III Power/Flow Map ...... * ......................................

46 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability O ption III Power/Flow M ap ...........................................................................

47* ' OPRM Operable, FWTR, 2923 MWt Figure 6: Stability O ption III Power/Flow M ap ...........................................................................

48 OPRM Inoperable, FWTR, 2923 MWt Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 6, Revision 0 APLHGR APRM ARTS BOC BSP BWROG CAVEX COLR CRWE DIVOM EFPD EOC EOC-RPT EOCLB EOFP EOOS F FHOOS FFTR FWTR HCOM ICF NOMENCLATURE Average Planar Linear Heat Generation Rate Average Power Range Monitor (Subsystem)

APRM/RBM Technical Specification Beginning of Cycle Backup Stability Protection BWR Owners Group Core Average Exposure Core Operating Limits Report Control Rod Withdrawal Error Delta CPR Over Initial MCPR Versus Oscillation Magnitude Effective Full Power Day End of Cycle End of Cycle Recirculation Pump Trip End of Cycle Licensing Basis End of Full Power Equipment Out of Service Flow (Total Core)Feedwater Heater Out of Service Final Feedwater Temperature Reduction Feedwater Temperature ReductionýHot Channel Oscillation Magnitude Increased Core Flow LCO LHGR LHGRss LHGRFAC LHGRFACf LHGRFACp LPRM MAPLHGR MAPLHGRss MAPFAC MAPFACf MAPFACp MAPFACSLO MCE MCPR MCPRf MCPRp Limiting Condition of Operation Linear Heat Generation Rate Steady-State Maximum Linear Heat Generation Rate Linear Heat Generation Rate Factor Flow-Dependent Linear Heat Generation Rate Factor Power-Dependent Linear Heat Generation Rate Factor Local Power Range Monitor (Subsystem)

Maximum Average Planar Linear Heat Generation Rate Steady-State Maximum Average Planar Linear Heat Generation Rate Maximum Average Planar Linear Heat Generation Rate Factor Flow-Dependent Maximum Average Planar Linear Heat Generation Rate Factor Power-Dependent Maximum Average Planar Linear Heat Generation Rate Factor Maximum Average Planar Linear Heat Generation Rate Factor when in SLO Maximum Core Exposure Minimum Critical Power Ratio Flow-Dependent Minimum Critical Power Ratio Power-Dependent Minimum Critical Power Ratio Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 7, Revision 0 NOMENCLATURE (continued)

MELLL Maximum Extended Load Line Limit MEOD Maximum Extended Operating Domain MSIVOOS Main Steam Isolation Valve Out of Service NEOC Near End of Cycle NFWT Nominal Feedwater Temperature NSS Nominal SCRAM Speed OLMCPR Operating Limit Minimum Critical Power Ratio OPRM Oscillation Power Range Monitor OOS Out of Service P Power (Total Core Thermal)PBDA Period Based Detection Algorithm PRNM Power Range Neutron Monitoring (System)RBM Rod Block Monitor (Subsystem)

RFWT Reduced Feedwater Temperature RPT Recirculation Pump Trip RTP Rated Thermal Power SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRV Safety Relief Valve SRVOOS Safety Relief Valve Out of Service STP Simulated Thermal Power TBV Turbine Bypass Valve TBVINS Turbine Bypass Valves In Service TBVOOS Turbine Bypass Valves Out of Service (all bypass valves OOS)TIP Traversing Incore Probe TLO Two Loop Operation TS Technical Specification TSSS Technical Specification SCRAM Speed Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design CaIc. No. 2B21-1293 Page 8, Revision 0 CAUTION References to COLR Figures or Tables should be made using titles only; Figure and Table numbers may change from cycle to cycle.I Introduction and Summary The Brunswick Unit 2, Cycle 19 Core Operating Limits Report (COLR) provides values for the core operation limits and setpoints required by Technical Specifications (TS) 5.6.5.a.Ite Requirement

~TS Table 3.3.1.1-1 Function 2f Average Power Range Monitors -OPRM Upscale.TS 3.3.1.1 LC Condio 1. Alternate method to detect and suppress thermal-hydraulic instability oscillations.

LCO Condition I TS Table 3.3.2.1-1 Rod Block Monitor- Upscale and Operability Requirements Function 1 TS 3.2.1 Average Planar Linear Heat Generation Rate (APLHGR).TS 3.2.2 Minimum Critical Power Ratio (MCPR).TS 3.2.3 Linear Heat Generation Rate (LHGR).TS LCO 3.4.1 APLHGR, MCPR and LHGR limits for SLO.TS LCO 3.7.6 APLHGR, MCPR and LHGR limits for an inoperable Main Turbine Bypass System.Core Operating Limits required to be documented in COLR:* APLHGR* MCPR TS 5.6.5.a* LHGR* PBDA setpoints 9 RBM allowable values and power range setpoints TS 5.6.5.b Analytical methods approved by the NRC for determining core operating limits.TS 5.6.5.c Core Operating Limits shall be determined such that all applicable limits of the safety analysis are met.TS 5.6.5.d The COLR shall be provided upon issuance for each cycle to the NRC.The core operating limits and setpoints presented in this COLR have been determined using NRC approved methodologies (References 7, 12, 14-30) in accordance with TS 5.6.5.b and are established such that all applicable limits of the plant safety analysis are met in accordance with TS 5.6.5.c.In addition to the TS required core operating limits and setpoints, this COLR also includes maps showing the allowable power/flow operating range including the Option III stability ranges.The generation of this COLR is documented in Reference 1 and is based on analysis results documented in References 2, 3, 4 and 5.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C1 9 Core Operating Limits Report Page 9, Revision 0 APLHGR Limits Steady-state MAPLHGRss limits are provided for ATRIUM-10 (Table 19) and GE14 (Tables 20 -22, 24-25). These steady-state MAPLHGRss limits must be modified as follows:* GE14 MAPLHGR limits have a core power and core flow dependency.

GE14 power-dependent MAPFACp multipliers (Table 26) and flow-dependent MAPFACf multipliers (Table 27) must be used to modify the steady-state MAPLHGRss limits (Tables 20 -22, 24 -25) for off-rated conditions." ATRIUM-10 MAPLHGR limits do not have a power or flow dependency.

However, in order to be consistent with the determination of GE14 MAPLHGR limits, power-dependent MAPFACp multipliers and flow-dependent MAPFACf multipliers with a constant value of 1.0 under all conditions have been assigned to ATRIUM-1 0." GE14 power-dependent MAPFACp multipliers include all allowed EOOS conditions as indicated in Table 26. See COLR section titled "Equipment Out of Service" for a list of analyzed EOOS conditions.

Care should be used when selecting the appropriate multiplier set." The applied MAPLHGR limit is dependent on the number of recirculation loops in operation.

The steady-state MAPLHGR limit must be modified by a MAPFACSLO multiplier when in SLO.MAPFACSLO has a fuel design dependency as shown below." All limits were established without assuming EOC-RPT. EOC-RPT is not required.The applied TLO and SLO MAPLHGR limits are determined as follows: MAPLHGR LimitTLO = MAPLHGRss x (MAPFACp, MAPFACf, 1.0)min MAPLHGR LimitSLO = MAPLHGRss x (MAPFACP, MAPFACf, MAPFACSLO)min where MAPFACSLO

= 0.85 for ATRIUM-10 fuel= 0.80 for GE14 fuel Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show two breakpoints at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0_), THEN select the most restrictive limit associated with the breakpoint.

Hand calculated results may not match a POWERPLEX calculation since normal monitoring of the APLHGR limits with POWERPLEX uses the complete set of lattices for each applicable fuel type provided in Reference 5 and as validated in Reference 9.MCPR Limits The MCPR limits presented in Tables 5 through 11 are based on the TLO SLMCPR listed in Technical Specification 2.1.1.2.* MCPR limits have a core power and core flow dependency.

Power-dependent MCPRp limits are presented in Tables 5 through 10 while flow-dependent MCPRf limits are presented in Table 11.* Power-dependent MCPRp limits are dependent on CAVEX, SCRAM insertion speed, EOOS, fuel design, number of operating recirculation loops (i.e., TLO or SLO) and core thermal power.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2821-1293 B2C19 Core Operating Limits Report Page 10, Revision 0 Values for the CAVEX breakpoints are provided in Table 4. See COLR section titled"Equipment Out of Service" for a list of analyzed EOOS conditions.

Care should be used when selecting the appropriate limits set.* The MCPR limits are established such that they bound all pressurization and non-pressurization events.* The power-dependent MCPRp limits (Tables 5-10) must be adjusted by an adder of 0.02 when in SLO.* All limits were established without assuming EOC-RPT. EOC-RPT is not required.The applied TLO and SLO MCPR limits are determined as follows: MCPR LimitTLO = (MCPRp, MCPRf)max MCPR LimitSLO = (MCPRP + 0.02, MCPRf)max Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show two breakpoints at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0), THEN select the most restrictive limit associated with the breakpoint.

LHGR Limits Steady-state LHGRss limits are provided for ATRIUM-10 (Table 12) and GE14 (Table 13). These steady-state LHGRss limits must be modified as follows:* ATRIUM-10 LHGR limits have a core power and core flow dependency.

ATRIUM-10 power-dependent LHGRFACp multipliers (Tables 14-17) and flow-dependent LHGRFACf multipliers (Table 18) must be used to modify the steady-state LHGRss limits (Table 12) for off-rated conditions.

• ATRIUM-10 power-dependent LHGRFACp multipliers are dependent on CAVEX, SCRAM insertion speed, EOOS, fuel design and core thermal power. Values for the CAVEX breakpoints are provided in Table 4. See COLR section titled "Equipment Out of Service" for a list of analyzed EOOS conditions.

Care should be used when selecting the appropriate multiplier set.* The original licensing basis for GE14 LHGR limits did not include a core power and core flow dependency.

However, in order to be consistent with the determination of ATRIUM-10 LHGR limits, power-dependent LHGRFACp multipliers and flow-dependent LHGRFACf multipliers with a constant value of 1.0 under all conditions have been assigned to,/GEl4.* GE14 LHGR limits are effectively monitored by GE14 APLHGR limits in accordance with the NRC approved methodology described in Reference 12." The applied LHGR limit is not dependent on the number of operating recirculation loops. No adjustment to the LHGR limit is necessary for SLO.* All limits were established without assuming EOC-RPT. EOC-RPT is not required.The applied LHGR limit is determined as follows: LHGR Limit = LHGRss x (LHGRFACp, LHGRFACf)min Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design CaIc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 11, Revision 0 Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show two breakpoints at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0), THEN select the most restrictive limit associated with the breakpoint.

PBDA Setpoints Brunswick Unit 2 has implemented BWROG Long Term Stability Solution Option III (OPRM) with the methodology described in Reference

6. Plant specific analysis incorporating the Option III hardware is described in Reference

13. Reload validation has been performed in accordance with Reference 7.The analysis was performed at 1 00%P assuming a two pump trip (2PT) and at 45%F assuming steady-state (SS) conditions at the highest rod line power (60.5%). The PBDA setpoints are set such that either the least limiting MCPRP limit or the least limiting MCPRf limit will provide adequate protection against violation of the SLMCPR during a postulated reactor instability.

Based on the MCPR limits presented in Tables 5 through 11, the required Amplitude Trip Setpoint (1.13) is set by the least limiting 100%P MCPRP limit (1.35) which has an associated Confirmation Count Setpoint (15). The PBDA setpoints shown in Table 3 are valid for any feedwater temperature.

Evaluations by General Electric (GE) have shown that the generic DIVOM curves specified in Reference 7 may not be conservative for current plant operating conditions for plants which have implemented Stability Option Ill. To address this issue, AREVA has performed calculations for the relative change in CPR as a function of the calculated HCOM. These calculations were performed with the RAMONA5-FA code in accordance with Reference

8. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in ACPR and the HCOM on a plant specific basis. The stability-based OLMCPRs are based upon using the most limiting ACPR calculated for a given oscillation magnitude or the generic value provided in Reference 7.In cases where the OPRM system is declared inoperable, Backup Stability Protection (BSP) in accordance with Reference 10 is provided.

Analyses have been performed to support operation with nominal feedwater temperature conditions and reduced feedwater temperature conditions (FHOOS and FFTR).The power/flow maps (Figures 1-6) were developed based on Reference 4 to facilitate operation under Stability Option III as implemented by Function 2.f of Table 3.3.1.1-1 and LCO Condition I of Technical Specification 3.3.1.1. All maps illustrate the region of the power/flow map above 25% RTP and below 60% drive flow (correlated to core flow) where the system is required to be enabled. The generation of these maps is documented in Reference 2.The maps supporting an operable OPRM (Figures 1, 3 and 5) show a Scram Avoidance Region, which is not a licensing requirement but is an operator aid to illustrate where the OPRM system may generate a scram to avoid an instability event. Note that the STP scram and rod block limits are defined in Technical Specifications, the Technical Requirements Manual, and Plant procedures, and are included in the COLR as an operator aid rather than a licensing requirement.

Figures 3 and 4 implement the corrective action for AR-217345217345which restricts reactor power to no more than 50% RTP when in SLO with OPRM operable or inoperable.

This operator aid is intended to mitigate a spurious OPRM trip signal which could result from APRM noise while operating at high power levels.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 12, Revision 0 RBM Setpoints The nominal trip setpoints and allowable values of the control rod withdrawal block instrumentation are presented in Table 1 and were determined to be consistent with the bases of the ARTS program (Reference 11). These setpoints will ensure the power-dependent MCPR limits will provide adequate protection against violation of the SLMCPR during a postulated CRWE event. Reference 3 revised these setpoints to reflect changes associated with the installation of the NUMAC PRNM system. RBM operability requirements, consistent with Notes (a) through (e) of Technical Specification Table 3.3.2.1-1, are provided in Table 2.Equipment Out-of-Service Brunswick Unit 2, Cycle 19 is analyzed for the following operating conditions with applicable MCPR, APLHGR and LHGR limits.* Base Case Operation* SLO* TBVOOS* FHOOS* Combined TBVOOS and FHOOS Base Case Operation as well as the above-listed EOOS assumes the following items OOS:* Any 1 inoperable SRV* 2 inoperable TBVs* Up to 40% of the TIP channels OOS* Up to 50% of the LPRMs OOS Please note that during FFTR/Coastdown, FHOOS is included in Base Case Operation and TBVOOS.Single Loop Operation Brunswick Unit 2, Cycle 19 may operate in SLO over the entire MEOD range with applicable MCPR, APLHGR and LHGR limits. The following must be considered when operating in SLO:* SLO is not permitted with FHOOS.* SLO is not permitted with TBVOOS.* SLO is not permitted with MSIVOOS.Various indicators on the Power/Flow Maps are provided not as operating limits but rather as a convenience for the operators.

The purposes for some of these indicators are as follows: The SLO Entry Rod Line is shown on the TLO maps to avoid regions of instability in the event of a pump trip.* A maximum core flow line is shown on the SLO maps to avoid vibration problems.* APRM STP Scram and Rod Block nominal trip setpoint limits are shown at the estimated core flow corresponding to the actual drive flow-based setpoints to indicate where the Operator may encounter these setpoints (See LCO 3.3.1.1, Reactor Protection System Instrumentation Function 2.b: Average Power Range Monitors Simulated Thermal Power -High Allowable Value).

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Caic. No. 2B21-1293 B2C19 Core Operating Limits Report Page 13, Revision 0 When in SLO, Figures 3 and 4 implement the corrective action for AR-217345217345which restricts reactor power to no more than 50% RTP with OPRM operable or inoperable.

This operator aid is intended to mitigate a spurious OPRM trip signal which could result from APRM noise while operating at high power levels..Inoperable Main Turbine Bypass System Brunswick Unit 2, Cycle 19 may operate with an inoperable Main Turbine Bypass System over the entire MEOD range and cycle with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. An operable Main Turbine Bypass System with only two inoperable bypass valves was assumed in the development of the Base Case Operation limits. Base Case Operation is synonymous with TBVINS. The following must be considered when operating with TBVOOS: " Three or more inoperable bypass valves renders the entire Main Turbine Bypass System inoperable requiring the use of TBVOOS limits. The TBVOOS analysis supports operation with all bypass valves inoperable.

• TBVOOS operation coincident with FHOOS is supported using the combined TBVOOS/FHOOS limits.* SLO is not permitted with TBVOOS.Feedwater Temperature Reduction Brunswick Unit 2, Cycle 19 may operate with RFWT over the entire MEOD range and cycle with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. NFWT is defined as the range of feedwater temperatures from NFWT to NFWT -1 0F. NFWT and its allowable variation were assumed in the development of the Base Case Operation limits. The FHOOS limits and FFTR/Coastdown limits were developed for a maximum feedwater temperature reduction of 110.3°F.The following must be considered when operating with RFWT:* Although the acronyms FWTR, FHOOS, RFWT and FFTR all involve reduced feedwater temperature, the use of FFTR is reserved for cycle energy extension using reduced feedwater temperature at and beyond a core average exposure of EOCLB using FFTR/Coastdown limits." Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10°F and reactor power > 30% RTP requires use of the FHOOS limits. Below 30% RTP, Base Case Operation limits bound FHOOS limits." Until a core average exposure of EOCLB is reached, implementation of the FFTR/Coastdown limits is not required even if coastdown begins early.* When operating with RFWT, the appropriate Stability Option III Power/Flow Maps (Figures 5 and 6) must be used." FHOOS operation coincident with TBVOOS is supported using the combined TBVOOS/FHOOS limits." SLO is not permitted with RFWT.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design CaIc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 14, Revision 0 References References designated with asterisks denote documents containing NRC approved methodologies listed in Brunswick Unit 2 Technical Specification 5.6.5.b.1. BNP Design Calculation 2B21-1293, "Preparation of the B2C19 Core Operating Limits Report," Revision 0.2. Design Calculation 0B21-1015, "BNP Power/Flow Maps," Revision 7.3. Design Calculation 2C51-0001, "Power Range Neutron Monitoring System Setpoint Uncertainty and Scaling Calculation (2-C51-APRM 1 through 4 Loops and 2-C51-RBM-A and B Loops)," Revision 3, May 2004.4. ANP-2771(P), "Brunswick Unit 2 Cycle 19 Reload Safety Analysis," Revision 0, January 2009.5. NEDC-31624P, "Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 2 Reload 17 Cycle 18," Supplement 2, Revision 10, January 2007.6. NEDO-31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology," November 1995.* 7. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Application," August 1996.8. BAW-10255PA, "Cycle Specific DIVOM Methodology Using the RAMONA5-FA Code," Revision 2, May 2008.9. GE Hitachi Nuclear Energy 0000-0091-3483-RO, "Evaluation of LOCA Analysis Effects from Installation of Adjustable Speed Drive and ECCS Performance Changes for'Brunswick," Revision 0, January 2009.10. OG02-0119-260 "Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy," July 17, 2002.11. NEDC-31654P, "Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant," February 1989.* 12. NEDE-2401 1-P-A, "GESTAR II -General Electric Standard Application for Reactor Fuel", and US Supplement, Revision 15, September 2005.13. GENE-C51-00251-00-01, "Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick 1 and 2," Revision 0, March 2001.* 14. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Revision 2, March 1984.* 15. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Revision 1, September 1986.* 16. EMF-85-74(P)

Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Revision 0, February 1998.* 17. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Revision 1, May 1995.* 18. XN-NF-80-19(P)(A)

Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis," March 1983.* 19. XN-NF-80-19(P)(A)

Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Revision 1, June 1986.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 15, Revision 0* 20. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Revision 0, October 1999.* 21. XN-NF-80-19(P)(A)

Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Revision 2, January 1987.* 22. XN-NF-84-105(P)(A)

Volume 1, "XCOBRA-T:

A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," February 1987.* 23. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," Revision 2, November 1990.* 24. ANF-913(P)(A)

Volume 1, "COTRANSA2:

A Computer Program for Boiling Water Reactor Transient Analyses," Revision 1, August 1990.* 25. ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Revision 3, September 2005.* 26. EMF-2209(P)(A), "SPCB Critical Power Correlation", Revision 2, September 2003.* 27. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Revision 0, August 2000.* 28. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model," Revision 0, May 2001.* 29. EMF-2292(P)(A), "ATRIUMTM-10:

Appendix K Spray Heat Transfer Coefficients," Revision 0, September 2000.* 30. EMF-CC-074(P)(A)

Volume 4, "BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2," Revision 0, August 2000.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 16, Revision 0 Table 1 RBM System Setpoints 1 LSetpoint a, d 11Trip Setpoint jFilicwable Value]Lower Power Setpoint (LPSP ) 27.7 < 29.0 Intermediate Power Setpoint (IPSP b 62.7 < 64.0 High Power Setpoint (HPSPb) 82.7 < 84.0 Low Trip Setpoint (LTSPc) < 114.1 < 114.6 Intermediate Trip Setpoint (ITSPc) < 108.3 < 108.8 High Trip Setpoint (HTSPc) < 104.5 < 105.0 RBM Time Delay (td2) < 2.0 seconds < 2.0 seconds a See Table 2 for RBM Operability Requirements.

b Se~tpoints in perc~ent of Rted Therma[ Power.c Setpoints relativeitoia full scale reading of 125. For example l < 114.1 mea ns< 1 14.1/125.0of full scale.4 d. Trip setpo nts and alowable values are based on a high power ainalytical S setpoin 0o 108%o(unfilteed)

.. .1 This table is referred to by Technical Specification 3.3.2.1 (Table 3.3.2.1-1) and 5.6.5.a.5.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 17, Revision 0 Table 2 RBM Operability Requirements 2 IF the following conditions are met, THEN RBM Not Required Operable Thermal Power (% rated) MCPR 1.53 TLO>29% and < 90% 1.55 SLO 9%1.55 SLO> 90% > 1.47 TLO 2 Requirements valid for all fuel designs, all SCRAM insertion times and all core average exposure ranges.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 18, Revision 0 Table 3 PBDA Setpoints 3 Amplitude Trip OLMCPR(SS)

OLMCPR(2PT) 1.05 1.20 1.18 1.06 1.22 1.20 1.07 1.24 1.21 1.08 1.26 1.23 1.09 1.28 1.25 1.10 1.30 1.27 i.11 1.32 1.29 1.12 1.34 1.31 1.13 1.36 1.33 1.14 1.38 1.35 1.15 1.40 1.38 AcceptanceCriteria QOff-rated OLMCPR @ Rated Power 45% Flow OLM~CPR~3 This table is referred to by Technical Specification 3.3.1.1 (Table 3.3.1.1-1) and 5.6.5.a.4.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 19, Revision 0 Table 4 Exposure Basis 4 for Brunswick Unit 2 Cycle 19 Transient Analysis Core Average Exposure Comments (MWd/MTU)Break point for exposure-dependent MCPRp 30,421 limits (NEOC)32,881 Design basis rod patterns to EOFP + 14 EFPD (EOCLB)34,776 End of reactivity for FFTR/Coastdown

-Maximum Core Exposure (MCE)4 The exposure basis for the defined break points is the core average exposure (CAVEX) values shown above regardless of the actual BOC CAVEX value of the As-Loaded Core.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 20, Revision 0 Table 5 Power-Dependent MCPRp Limits 5 NSS Insertion Times BOC to < NEOC (EOC-RPT not required)EOOS Power ATRIUM-10 GE14'Condition

(% rated) MCPRP MCPRp 100.0 1.35 1.37 90.0 1.38 1.40 50.0 1.48 1.48 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.39 1.40 90.0 1.42 1.43 50.0 1.57 1.59> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 100.0 1.36 1.38 90.0 1.39 1.41 50.0 1.51 1.52> 65%F < 65%F > 65%F < 65%F 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.40 1.41 90.0 1.43 1.45 TBVOOS 50.0 1.61 1.63 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.94 1.73 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Table 6 Power-Dependent.

MCPRp Limits 6 TSSS Insertion Times BOC to < NEOC (EOC-RPT not required)Design Calc. No. 2B21-1293 Page 21, Revision 0 EOOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.45 1.47 90.0 1.46 1.48 50.0 1.50 1.52 Base Case > 65%F

• 65%F > 65%F < 65%F Operation 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.47 1.49 90.0 1.49 1.52 50.0 1.63 1.65> 65%F < 65%F > 65%F < 65%F 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 100.0 1.46 1.48 90.0 1.47 1.49 50.0 1.53 1.54> 65%F < 65%F > 65%F < 65%F 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.48 1.50 90.0 1.50 1.53 TBVOOS 50.0 1.67 1.68 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.97 1.75 2.03 1.79 26.0 2.29 2.03 2.25 .2.04.26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 6 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Table 7 Power-Dependent MCPRp Limits 7 NSS Insertion Times BOC to < EOCLB (EOC-RPT not required)Design Calc. No. 2B21-1293 Page 22, Revision 0 EOOS Power ATRIUM-10 GEI4 Condition

(% rated) MCPRP MCPRP 100.0 1.43 1.44 90.0 1.45 1.45 50.0 1.53 1.53 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 1.95 1.74 2.00 .1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.45 1.46 90.0 1.48 1.48 50.0 1.59 1.59> 65%F < 65%F > 65%F < 65%F 50.0 1.95 1.74 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 100.0 1.44 1.45 90.0 1.46 1.46 50.0 1.53 1.53 FHOOS > 65%F < 65%F > 65%F < 65%F 50.0 1.95 1.74 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.46 1.50 90.0 1.49 1.52 TBVOOS 50.0 1.62 1.63 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.95 1.74 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Caic. No. 2B21-1293 Page 23, Revision 0 Table 8 Power-Dependent MCPRp TSSS Insertion Times BOC to < EOCLB (EOC-RPT not required)Limits 8 EOOS Power ATRIUM-10 GE14 Condition

(% rated) MCPRp MCPRp 100.0 1.48 1.49 90.0 1.49 1.49 50.0 1.53 1.53 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 1.99 1.77 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.50 1.51 90.0 1.51 1.53 50.0 1.63 1.65> 65%F <65%F > 65%F <65%F 50.0 1.99 1.77 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.95 2.64 2.95 2.67 23.0 3.09 2.87 3.10 2.90 100.0 1.49 1.50 90.0 1.50 1.50 50.0 1.53 1.54> 65%F < 65%F > 65%F < 65%F 50.0 1.99 1.77 2.03 1.79 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.51 1.59 90.0 1.52 1.61 TBVOOS 50.0 1.67 1.72 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.99 1.77 2.07 1.83 26.0 2.29 2.03 2.29 2.08 26.0 2.95 2.64 2.95 2.67 23.0 3.09 2.87 3.10 2.90 8 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design CaIc. No. 2B21-1293 Page 24, Revision 0 Table 9 Power-Dependent MCPRp Limits 9 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)(EOC-RPT not required)EOOS 10 Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.47 1.46 90.0 1.48 1.47 Base Case 50.0 1.56 1.55 Operation

> 65%F < 65%F > 65%F < 65%F 50.0 1.96 1.75 2.00 1.77 (includes FHOOS) 26.0 2.27 2.02 2.23 2.03 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.47 1.50 90.0 1.49 1.52 50.0 1.62 1.63 TBVOOS > 65%F < 65%F > 65%F < 65%F 50.0 1.96 1.75 2.00 1.77 26.0 2.27 2.02 2.23 2.03 26.0 2.95 2.64 2.91 2.63 23.0 3.09 2.87 3.06 2.86 9 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.10 FFTR/FHOOS included in Base Case Operation and TBVOOS.

f Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 25, Revision 0 Table 10 Power-Dependent MCPRp Limits 1 1 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)(EOC-RPT not required)EOOS 12 Power ATRIUM-10 GE14 Condition

(% rated) MCPRP MCPRP 100.0 1.50 1.50 90.0 1.50 1.50 Base Case 50.0 1.56 1.55 Operation

> 65%F <65%F > 65%F < 65%F 50.0 2.02 1.81 2.04 1.80 (includes FHOOS) 26.0 2.29 2.03 2.25 2.04 26.0 2.30 2.04 2.26 2.05 23.0 2.35 2.12 2.30 2.12 100.0 1.51 1.59 90.0 1.52 1.61 50.0 1.67 1.72 TBVOOS > 65%F < 65%F > 65%F < 65%F 50.0 2.02 1.81 2.07 1.83 26.0 2.29 2.03 2.29 2.08 26.0 2.95 2.64 2.95 2.67' 23.0 3.09 2.87 3.10 2.90 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.12 FFTR/FHOOS included in Base Case Operation and TBVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 26, Revision 0 Table 11 Flow-Dependent MCPRf Limits 1 3'1 4 Core Flow ATRIUM-10/GE14

(% of rated) MCPRf 0'0 1.65 31.0 1.65 100.0 1.20 107.0 1.20 13 Limits valid for all SCRAM insertion times and all core average exposure ranges.14 EOC-RPT is not applicable for this limit.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 27, Revision 0 Table 12 ATRIUM-10 Steady-State LHGRss Limits Peak ATRIUM-10 Pellet Exposure LHGR (GWd/MTU) (kW/ft)0.0 13.4 18.9 13.4 74.4 7.1 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 28, Revision 0 Table 13 GE14 Steady-State LHGRss Limits 1 5' 1 6 Peak GE14 Pellet Exposure LHGR (GWd/MTU) (kW/ft)All 13.4'5 GE14 LHGR limits are effectively monitored by GE14 MAPLHGR limits in accordance with the NRC approved methodology described in Reference 12.16 GE14 LHGR limits do not have a power or flow dependency.

Thus, the GE14 LHGRFACp and the LHGRFACf multipliers have a constant value of 1.0 under all conditions.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 29, Revision 0 Table 14 ATRIUM-10 Power-Dependent LHGRFACP Multipliers 1 7 NSS Insertion Times BOC to < EOCLB (EOC-RPT not required)EOOS Power ATRIUM-10 Condition

(% rated) LHGRFACp 100.0 1.00 90.0 1.00 50.0 0.90 Base Case. > 65%F < 65%F Operation 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.95 90.0 0.94 50.0 0.86> 65%F <65%F TBVOOS 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.45 0.51 23.0 0.43 0.47 100.0 1.00 90.0 1.00 50.0 0.90> 65%F <65%F 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.95 90.0 0.94 50.0 0.85 TBVOOS> 65%F < 65%F and and 50.0 0.69 0.78 26.0 0.60 0.65 26.0 0.45 0.51 23.0 0.43 0.47 17 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 30, Revision 0 Table 15 ATRIUM-10 Power-Dependent LHGRFACp Multipliers 1 8 TSSS Insertion Times BOC to < EOCLB (EOC-RPT not required)EOOS Power ATRIUM-10 Condition

(% rated) LHGRFACP 100.0 1.00 90.0 0.95 50.0 0.87 Base Case > 65%F < 65%F Operation 50.0 0.67 0.77 26.0 0.60 0.64 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.93 90.0 0.91 50.0 0.86> 65%F < 65%F TBVOOS 50.0 0.67 0.77 26.0 0.60 0.64 26.0 0.45 0.51 23.0 0.43 0.47 100.0 1.00 90.0 0.95 50.0 0.87> 65%F < 65%F 50.0 0.67 0.77 26.0 0.60 0.64 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.93 90.0 0.91 TBVOOS 50.0 0.84 ad> 65%F < 65%F and 50.0 0.67 0.77 FHOOS 26.0 0.60 0.64 26.0 0.45 0.51 23.0 0.43 0.47 18 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 31, Revision 0 Table 16 ATRIUM-10 Power-Dependent LHGRFACp Multipliers 1 9 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)(EOC-RPT not required)EOOS 2 0 Power ATRIUM-10 Condition

(% rated) LHGRFACP 100.0 1.00 90.0 1.00 Base Case 50.0 0.86 Operation

> 65%F < 65%F 50.0 0.69 0.77 (includes FHOOS) 26.0 0.60 0.65 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.93 90.0 0.91 50.0 0.85 TBVOOS > 65%F < 65%F (includes FHOOS) 50.0 0.69 0.77 26.0 0.60 0.65 26.0 0.45 0.51 23.0 0.43 0.47 19 Limits sup'port operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

21 FFTR/FHOOS included in Base Case Operation and TBVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 32, Revision 0 Table 17 ATRIUM-10 Power-Dependent LHGRFACP Multipliers 2 1 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)(EOC-RPT not required)EOOS 2 2 Power ATRIUM-10 Condition

(% rated) LHGRFACP 100.0 1.00 90.0 0.94 Base Case 50.0 0.86 Operation

> 65%F < 65%F 50.0 0.67 0.76 (includes FHOOS) 26.0 0.60 0.64 26.0 0.59 0.63 23.0 0.58 0.62 100.0 0.91 90.0 0.90 50.0 0.82 TBVOOS > 65%F -< 65%F (includes FHOOS) 50.0 0.67 0.76 26.0. 0.60 0.64 26.0 0.45 0.51 23.0 0.43 0.47 21 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBVs, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

22 FFTR/FHOOS included in Base Case Operation and TBVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 33, Revision 0 Table 18 ATRIUM-10 Flow-Dependent LHGRFACf Multipliers 2 3'2 4 Core Flow (% of rated) LHGRFACf 0.0 0.90 31.0 0.90 50.0 1.00 107.0 1.00 23 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.24 EOC-RPT is not applicable for this limit.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B211-1293 Page 34, Revision 0 Table 19 ATRIUM-10 Steady-State MAPLHGRss Limits 2 5' 2 6 Average Planar Exposure (GWd/MTU)

MAPLHGR (kW/ft)0.0 12.5 15.0 12.5 67.0 7.3 2' ATRIUM-1 0 MAPLHGR limits do not have a power or flow dependency.

Thus, tIhe ATRIUM-1 0 MAPFACp and the MAPFACf multipliers have a constant value of 1.0 under all conditions.

2' ATRIUM-10 MAPLHGR limits must be adjusted by a 0.85 multiplier when in SLO. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 35, Revision 0 Table 20 GEl4 Steady-State MAPLHGRss Limits 2 7 GEl 4-P1ODNAB425-3G7.0/14G6.0/1G2.0-10OT-150-T-2574 (only)Average Planar Exposure Limit (GWd/MTU) (kW/ft)0.00 9.37 0.22 9.44 1.10 9.55 2.20 9.70 3.31 9.85 4.41 10.01 5.51 10.16 6.61 10.30 7.72 10.45 8.82 10.60 9.92 10.70 11.02 10.79 12.13 10.87 13.23 10.87 14.33 10.85 15.43 10.84 16.00 10.84 16.53 10.84 18.74 10.82 21.09 10.72 22.05 10.68 27.56 10.26 33.07 9.82 38.58 9.33 44.09 8.83 49.60 8.35 55.12 7.85 60.63 5.72 62.46 4.90 27 The GE14 MAPLHGR limits presented in this COLR are a composite set of limits based on the most limiting MAPLHGR limits from each lattice.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 36, Revision 0 Table 21 GE14 Steady-State MAPLHGRss Limits 2 8 GE14-P1ODNAB439-12G6.0-10OT-150-T-2575 (only)Average Planar Exposure Limit (GWd/MTU) (kW/ft)0.00 9.68 0.22 9.72 1.10 9.79 2.20 9.89 3.31 9.99 4.41 10.09 5.51 10.20 6.61 10.31 7.72 10.43 8.82 10.55 9.92 10.67 11.02 10.79 12.13 10.92 13.23 10.93 14.33 10.92 15.43 10.90 16.00 10.89 16.53 10.88 18.74 10.81 21.09 10.66 22.05 10.60 27.56 10.18 33.07 9.76 38.58 9.32 44.09 8.87 49.60 8.37 55.12 7.83 60.63 5.54 62.06 4.88 28 The GE14 MAPLHGR limits presented in this COLR are a composite set of limits based on the most limiting MAPLHGR limits from each lattice.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 37, Revision 0 Table 22 GEl4 Steady-State MAPLHGRss Limits 2 9 GEl 4-P1ODNAB413-16GZ-10OT-150-T-2660 (only)Average Planar Exposure Limit (GWd/MTU) (kW/ft)0.00 9.57 0.22 9.60 1.10 9.66 2.20 9.77 3.31 9.92 4.41 10.08 5.51 10.26 6.61 10.44 7.72 10.59 8.82 10.74 9.92 10.87 11.02 11.00 12.13 11.12 13.23 11.15 14.33 11.16 15.43 11.16 16.00 11.16 16.53 11.16 18.74 11.13 21.09 11.02 22.05 10.98 27.56 10.57 33.07 10.15 38.58 9.65 44.09 9.12 49.60 8.59 55.12 8.04 60.63 6.48 63.50 5.18 64.16 4.88 29 The GE14 MAPLHGR limits presented in this COLR are a composite set of limits based on the most limiting MAPLHGR limits from each lattice.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 38, Revision 0 Table 23[Not Used]

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 39, Revision 0 Table 24 GE14 Steady-State MAPLHGRss Limits 3 0 GE14-P1ODNAB407-16GZ-10OT-150-T-2853 (only)Average Planar Exposure Limit (GWd/MTU) (kW/ft)0.00 9.47 1.10 9.71 2.20 9.87 3.31 10.02 4.41 10.17 5.51 10.31 6.61 10.45 7.72 10.57 8.82 10.70 9.92 10.82 11.02 10.93 12.13 11.05 13.23 11.06 14.33 11.05 15.43 11.05 16.00 11.05 16.53 11.05 18.74 11.02 21.09 10.91 22.05 10.87 27.56 10.49 33.07 10.08 38.58 9.58 44.09 9.07 49.60 8.54 55.12 7.99 60.63 6.34 63.50 5.03 63.84 4.88 30 The GE14 MAPLHGR limits presented in this COLR are a composite set of limits based on the most limiting MAPLHGR limits from each lattice.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis Design Calc. No. 2B21-1293 B2C19 Core Operating Limits Report Page 40, Revision 0 Table 25 GE14 Steady-State MAPLHGRss Limits 3 1 GE14-P1ODNAB425-18.GZ-100T-150-T-2854 (only)Average Planar Exposure Limit (GWd/Mt) (kW/ft)0.00 8.85 1.10 9.01 2.20 9.14 3.31 9.27 4.41 9.41 5.51 9.54 6.61 9.68 7.72 9.82 8.82 9.95 9.92 10.09 11.02 10.23 12.13 10.27 13.23 10.29 14.33 10.31 15.43 10.34 16.00 10.35 16.53 10.37 18.74 10.44 21.09 10.47 22.05 10.48 27.56 10.28 33.07 9.84 38.58 9.38 44.09 8.91 49.60 8.41 55.12 7.88 60.63 5.70 62.23 4.96 31 The GE14 MAPLHGR limits presented in this COLR are a composite set of limits based on the most limiting MAPLHGR limits from each lattice.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 41, Revision 0 Table 26 GEl4 Power-Dependent MAPFACp Multipliers 3 2'3 3 (EOC-RPT not required)EOOS Power GE14 Condition

(% rated) MAPFACp 100.0 1.00 50.0 0.73 Base Case > 65%F < 65%F and all supported 50.0 0.64 0.73 EOOS Conditions 26.0 0.56 0.61 26.0 0.43 0.49 23.0 0.41 0.45 32 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.33 The GE14 power-dependent and flow-dependent multipliers are capped at 0.80 when in SLO. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Design Calc. No. 2B21-1293 Page 42, Revision 0 Table 27 GEl4 Flow-Dependent MAPFACf Multipliers 3 4'3 5 Core Flow GE14 (% rated) MAPFACf 0.0 0.56 31.0 0.56 80.0 1.00 107.0 1.00 34 The GE14 power-dependent and flow-dependent multipliers are capped at 0.80 when in SLO. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.35 EOC-RPT is not applicable for this limit.

Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Figure 1 Stability Option III Power/Flow Map OPRM Operable, Two Loop Operation, 2923 MWt Design Calc. No. 2B21-1293 Page 43, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 1 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow% Mlsb lsh 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow

Reference:

0B21-1015, Revision 7 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Figure 2 Stability Option Ill Power/Flow Map OPRM Inoperable, Two Loop Operation, 2923 MWt Design Calc. No. 2B21-1293 Page 44, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum ( MELLL) (ICF)Core Core Power Flow Flow% Mibs/lir Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow

Reference:

0121-1015, Revision 7 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Figure 3 Stability Option III Power/Flow Map OPRM Operable, Single Loop Operation, 2923 MWt Design Calc. No. 2B21-1293 Page 45, Revision 0 ---.This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum[MELLL) (ICF)Core Core Power Flow Flow/. Mlbs/hr M b r 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99. 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 ,46.2 53.9 61.6 69.3 77.0 84.7- 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow

Reference:

0B21-1015, Revision 7 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Figure 4 Stability Option III Power/Flow Map OPRM Inoperable, Single Loop Operation, 2923 MWt Design Calc. No. 2B21-1293 Page 46, Revision 0 This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Cure Power Flow Flow% Mlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow

Reference:

0B21-1015, Revision 7 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Figure 5 Stability Option Ill Power/Flow Map OPRM Operable, FWTR, 2923 MWt Design Calc. No. 2B21-1293 Page 47, Revision 0 This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 0.50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow% Mlbs/lir Mlbs/h r 100 76.19 80.47 99 75.04 80.47.... 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80:47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow

Reference:

0B21-1015, Revision 7 Progress Energy Nuclear Fuels Mgmt. and Safety Analysis B2C19 Core Operating Limits Report Figure 6 Stability Option III Power/Flow Map OPRM Inoperable, FWTR, 2923 MWt Design Cale. No. 2B21-1293 Page 48, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow% Mlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 MIlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow

Reference:

0B21-1015, Revision 7 BSEP 09-0034 Enclosure 3 AREVA Affidavit Regarding Withholding ANP-2729(P), Revision 0, from Public Disclosure AFFIDAVIT STATE OF VIRGINIA )) ss.CITY OF LYNCHBURG

)1. My name is Mark J. Burzynski.

I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information provided in support of a'Progress Energy License Amendment Request for Brunswick Unit 2 regarding use of AREVA NP ATRIUMTm-10 fuel. The following AREVA NP document is provided and referred to herein as the "Document."* AREVA NP document ANP-2729(P), Revision 0, Brunswick Unit 2 Thermal-Hydraulic Design Report for ATRIUMTM-10 Fuel Assemblies, dated July 2008 Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information".

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c) and 6(d) above.7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this Z -3 dayof _________

__,2008.Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/10 Reg. # 7079129 CommOnwalth of Virginia My C o7939 O9 ory Cmmlslofl fxpimi Oct 31. 2010r