BSEP 17-0032, Cycle 23 Core Operating Limits Report (COLR)

Cycle 23 Core Operating Limits Report (COLR)

ML17100A840
Person / Time
Site:	Brunswick
Issue date:	04/09/2017
From:	Mcpherson M Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 17-0032
Download: ML17100A840 (44)

Text

Brunswick Nuclear Plant

~~ DUKE P.O. Box 10429

~ ENERGY Southport, NC 28461 APR 0 9 2017 10 CFR 50.4 Serial: BSEP 17-0032 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit No. 2 Renewed Facility Operating License No. DPR-62 Docket Nos. 50-324 Unit 2 Cycle 23 Core Operating Limits Report (COLR)

Reference:

1. Letter from Annette H. Pope (Duke Energy) to NRC Document Control Desk, Unit 2 Cycle 22 Core Operating Limits Report (COLR), dated March 23, 2015, ADAMS Accession Number ML15091A406 Ladies and Gentlemen:

Enclosed is a copy of the Core Operating Limits Report (COLR) for Brunswick Steam Electric Plant (BSEP), Unit 2 Cycle 23 operation. Duke Energy Progress, LLC, is providing the enclosed COLR in accordance with Brunswick Unit 2 Technical Specification 5.6.5.d. The enclosed COLR supersedes the report previously submitted by letter dated March 23, 2015 (i.e., Reference 1).

This letter and the enclosed COLR do not contain any regulatory commitments.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

Sincerely, Mark McPherson Director - Organizational Effectiveness (Acting)

Brunswick Steam Electric Plant

U.S. Nuclear Regulatory Commission Page 2 of 2 WRM/wrm

Enclosure:

Brunswick Unit 2, Cycle 23 Core Operating Limits Report cc (with enclosure):

U.S. Nuclear Regulatory Commission, Region II ATTN: Ms. Catherine Haney, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 Andrew.Hon@nrc.gov U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission (Electronic Copy Only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net

BSEP 17-0032 Enclosure Brunswick Unit 2, Cycle 23 Core Operating Limits Report

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Cale. No. 2821-1325 82C23 Core Operating Limits Report, BNEl-0400-0005 Page 1, Revision 0 BRUNSWICK UNIT 2, CYCLE 23 CORE OPERATING LIMITS REPORT April 2017 Prepared By:

Cody J. Gilbert Brunswick Nuclear Design Verified By:

Ckb~4/¥~17 Brunswick Nuclear Design BNEl-0400-0005 Revision 0 I Attachment 1 I Page 2 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 2, Revision 0 LIST OF EFFECTIVE PAGES Page(s) Revision 1- 41 0 This document consists of 41 total pages.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 3 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 3, Revision 0 TABLE OF CONTENTS Subject Page Cover ............................................................................................................................................... 1 List of Effective Pages...................................................................................................................... 2 Table of Contents............................................................................................................................. 3 List of Tables.................................................................................................................................... 4 List of Figures .................................................................................................................................. 5 Nomenclature................................................................................................................................... 6 Introduction and Summary ............................................................................................................... 8 APLHGR Limits ................................................................................................................................ 9 MCPR Limits .................................................................................................................................... 9 LHGR Limits................................................................................................................................... 10 PBDA Setpoints ............................................................................................................................. 10 RBM Setpoints ............................................................................................................................... 11 Equipment Out-of-Service .............................................................................................................. 11 Single Loop Operation.................................................................................................................... 12 Inoperable Main Turbine Bypass System ....................................................................................... 12 Feedwater Temperature Reduction ................................................................................................ 13 Pressure Regulator Out-of-Service................................................................................................. 13 References..................................................................................................................................... 14 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 4 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 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 Page Table 1: RBM System Setpoints ................................................................................................. 16 Table 2: RBM Operability Requirements..................................................................................... 17 Table 3: PBDA Setpoints ............................................................................................................ 18 Table 4: Exposure Basis for Brunswick Unit 2 Cycle 23 Transient Analysis ................................ 19 Table 5: Power-Dependent MCPR p Limits.................................................................................. 20 NSS Insertion Times - BOC to < EOCLB Table 6: Power-Dependent MCPR p Limits.................................................................................. 22 TSSS Insertion Times - BOC to < EOCLB Table 7: Power-Dependent MCPR p Limits.................................................................................. 24 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 8: Power-Dependent MCPR p Limits.................................................................................. 25 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 9: Flow-Dependent MCPR f Limits..................................................................................... 26 Table 10: AREVA Fuel Steady-State LHGR SS Limits.................................................................... 27 Table 11: AREVA Fuel Power-Dependent LHGRFAC p Multipliers................................................ 28 NSS Insertion Times - BOC to < EOCLB Table 12: AREVA Fuel Power-Dependent LHGRFAC p Multipliers................................................ 30 TSSS Insertion Times - BOC to < EOCLB Table 13: AREVA Fuel Power-Dependent LHGRFAC p Multipliers................................................ 32 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 14: AREVA Fuel Power-Dependent LHGRFAC p Multipliers................................................ 33 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 15: AREVA Fuel Flow-Dependent LHGRFAC f Multipliers ................................................... 34 Table 16: AREVA Fuel Steady-State MAPLHGR SS Limits ............................................................ 35 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 5 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 5, 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 FIGURES Figure Title or Description Page Figure 1: Stability Option III Power/Flow Map .............................................................................. 36 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability Option III Power/Flow Map .............................................................................. 37 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability Option III Power/Flow Map .............................................................................. 38 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability Option III Power/Flow Map .............................................................................. 39 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability Option III Power/Flow Map .............................................................................. 40 OPRM Operable, FWTR, 2923 MWt Figure 6: Stability Option III Power/Flow Map .............................................................................. 41 OPRM Inoperable, FWTR, 2923 MWt BNEI-0400-0005 Revision 0 l Attachment 1 l Page 6 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 6, Revision 0 NOMENCLATURE 2PT Two Recirculation Pump Trip APLHGR Average Planar Linear Heat Generation Rate APRM Average Power Range Monitor (Subsystem)

ARTS APRM/RBM Technical Specification BOC Beginning of Cycle BSP Backup Stability Protection BWROG BWR Owners Group CAVEX Core Average Exposure COLR Core Operating Limits Report CRWE Control Rod Withdrawal Error DIVOM Delta CPR Over Initial MCPR Versus Oscillation Magnitude EFPD Effective Full Power Day EOC End of Cycle EOCLB End of Cycle Licensing Basis EOFP End of Full Power EOOS Equipment Out-of-Service F Flow (Total Core)

FHOOS Feedwater Heater Out-of-Service FFTR Final Feedwater Temperature Reduction FWTR Feedwater Temperature Reduction GE General Electric HCOM Hot Channel Oscillation Magnitude HPSP High Power Set Point HTSP High Trip Set Point ICF Increased Core Flow IPSP Intermediate Power Set Point ITSP Intermediate Trip Set Point LCO Limiting Condition of Operation LHGR Linear Heat Generation Rate LHGR SS Steady-State Maximum Linear Heat Generation Rate LHGRFAC Linear Heat Generation Rate Factor LHGRFAC f Flow-Dependent Linear Heat Generation Rate Factor LHGRFAC p Power-Dependent Linear Heat Generation Rate Factor LPRM Local Power Range Monitor (Subsystem)

LPSP Low Power Set Point LTA Lead Test Assembly LTSP Low Trip Set Point MAPLHGR Maximum Average Planar Linear Heat Generation Rate MAPLHGR SS Steady-State Maximum Average Planar Linear Heat Generation Rate MAPFAC Maximum Average Planar Linear Heat Generation Rate Factor BNEI-0400-0005 Revision 0 l Attachment 1 l Page 7 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 7, Revision 0 NOMENCLATURE (continued)

MAPFAC f Flow-Dependent Maximum Average Planar Linear Heat Generation Rate Factor MAPFAC p Power-Dependent Maximum Average Planar Linear Heat Generation Rate Factor MAPFAC SLO Maximum Average Planar Linear Heat Generation Rate Factor when in SLO MCE Maximum Core Exposure MCPR Minimum Critical Power Ratio MCPR f Flow-Dependent Minimum Critical Power Ratio MCPR p Power-Dependent Minimum Critical Power Ratio 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 NRC Nuclear Regulatory Commission 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)

PROOS Pressure Regulator Out-of-Service 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 SS Steady-State 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 8 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 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.

Introduction and Summary The Brunswick Unit 2, Cycle 23 COLR provides values for the core operation limits and setpoints required by Technical Specifications (TS) 5.6.5.a.

NRC Required Core Approved Operating Limit Related TS Items Methodology (TS 5.6.5.a )

(TS 5.6.5.b)

1. APLHGR for TS 3.2.1. 1, 2, 6, 7,16,  TS 3.2.1 LCO (APLHGR) 17  TS 3.4.1 LCO (Recirculation loops operating)

 TS 3.7.6 LCO (Main Turbine Bypass out-of-service)

2. MCPR for TS 3.2.2. 1, 2, 6, 7, 8, 9,  TS 3.2.2 LCO (MCPR) 10, 11, 12, 13,  TS 3.4.1 LCO (Recirculation loops 14, 21 operating)

 TS 3.7.6 LCO (Main Turbine bypass out-of-service)

3. LHGR for TS 3.2.3. 2, 3, 4, 5, 6, 7,  TS 3.2.3 LCO (LHGR) 8, 9, 10, 12,  TS 3.4.1 LCO (Recirculation loops 13, 20 operating)

 TS 3.7.6 LCO (Main Turbine bypass out-of-service)

4. PBDA setpoint for 8, 14, 18, 19,  TS Table 3.3.1.1-1, Function 2.f Function 2.f, APRM - OPRM 21 (APRM - OPRM Upscale)

Upscale, for TS 3.3.1.1.

 TS 3.3.1.1, Condition I (Alternate instability detection and suppression)

5. The Allowable Values and 6, 8  TS Table 3.3.2.1-1, Function 1 (RBM power range setpoints for Rod upscale and operability requirements)

Block Monitor Upscale Functions for TS 3.3.2.1.

The required core operating limits and setpoints listed in TS 5.6.5.a are presented in the COLR, have been determined using NRC approved methodologies (COLR References 1 through 21) in accordance with TS 5.6.5.b, have considered all fuel types utilized in B2C23, 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 30 and is based on analysis results documented in References 27-29.

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 9, Revision 0 APLHGR Limits Steady-state MAPLHGR SS limits are provided for AREVA Fuel (Table 16). These steady-state MAPLHGR SS limits must be modified as follows:

x AREVA Fuel MAPLHGR limits do not have a power, flow, or EOOS dependency.

x The applied MAPLHGR limit is dependent on the number of recirculation loops in operation. The steady-state MAPLHGR limit must be modified by a MAPFAC SLO multiplier when in SLO.

MAPFAC SLO has a fuel design dependency as shown below.

The applied TLO and SLO MAPLHGR limits are determined as follows:

MAPLHGR Limit TLO = MAPLHGR SS MAPLHGR Limit SLO = MAPLHGR SS x MAPFAC SLO where MAPFAC SLO = 0.80 for ATRIUM 10XM and ATRIUM 11 fuel Linear interpolation should be used to determine intermediate values between the values listed in the table.

MCPR Limits The MCPR limits presented in Tables 5 through 9 are based on the TLO and SLO SLMCPRs listed in Technical Specification 2.1.1.2 of 1.07 and 1.09, respectively.

x MCPR limits have a core power and core flow dependency. Power-dependent MCPR p limits are presented in Tables 5 through 8 while flow-dependent MCPR f limits are presented in Table 9.

x Power-dependent MCPR P limits are dependent on CAVEX, SCRAM insertion speed, EOOS, fuel design, number of operating recirculation loops (i.e., TLO or SLO), core flow 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 limits set.

x The MCPR limits are established such that they bound all pressurization and non-pressurization events.

x The power-dependent MCPR p limits (Tables 5-8) must be adjusted by an adder of +0.02 when in SLO.

The applied TLO and SLO MCPR limits are determined as follows:

MCPR Limit TLO = (MCPR p , MCPR f ) max MCPR Limit SLO = (MCPR p + 0.02, MCPR f ) max Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show step changes at 26.0%P and 50.0%P. A subset of EOOS limits show an additional step change at 80%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0, 50.0 or 80.0), THEN select the most restrictive limit associated with the breakpoint.

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 10, Revision 0 LHGR Limits Steady-state LHGR SS limits are provided for AREVA Fuel (Table 10). These steady-state LHGR SS limits must be modified as follows:

x AREVA Fuel LHGR limits have a core power and core flow dependency. AREVA Fuel power-dependent LHGRFAC p multipliers (Tables 11-14) and flow-dependent LHGRFAC f multipliers (Table 15) must be used to modify the steady-state LHGR SS limits (Table 10) for off-rated conditions.

x AREVA Fuel power-dependent LHGRFAC p multipliers are dependent on CAVEX, SCRAM insertion speed, EOOS, fuel design, core flow 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.

x The applied LHGR limit is not dependent on the number of operating recirculation loops. No adjustment to the LHGR limit is necessary for SLO.

The applied LHGR limit is determined as follows:

LHGR Limit = LHGR SS x (LHGRFAC p , LHGRFAC f ) min Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show step changes 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 23. Plant specific analysis incorporating the Option III hardware is described in Reference 24. Reload validation has been performed in accordance with Reference 19. The analysis was performed at 100%P assuming a two pump trip (2PT) and at 45%F assuming steady-state (SS) conditions at the highest rod line power (60.6%). The PBDA setpoints are set such that either the least limiting MCPR p limit or the least limiting MCPR f 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 9, the required Amplitude Trip Setpoint (1.10) is set by the least limiting 100%P MCPR p limit (1.34) with an allowance for conservative margin, which has an associated Confirmation Count Setpoint (13). The PBDA setpoints shown in Table 3 are valid for any feedwater temperature.

Evaluations by GE have shown that the generic DIVOM curves specified in Reference 19 may not be conservative for current plant operating conditions for plants which have implemented Stability Option III.

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 26. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in CPR and the HCOM on a plant specific basis. The stability-based OLMCPRs are based upon using the most limiting 'CPR calculated for a given oscillation magnitude or the generic value provided in Reference 19.

In cases where the OPRM system is declared inoperable, Backup Stability Protection (BSP) in accordance with Reference 25 is provided. Analyses have been performed to support operation with nominal feedwater temperature conditions and reduced feedwater temperature conditions (FHOOS and FFTR).

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 11, Revision 0 The power/flow maps (Figures 1-6) were validated for B2C23 based on Reference 29 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. The generation of these maps is documented in Reference 28. 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. Figures 1-6 were included in the COLR as an operator aid and not a licensing requirement. Figures 5 and 6 are the power/flow maps for use in FWTR.

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/or 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.

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 22). These setpoints will ensure the power-dependent MCPR limits will provide adequate protection against violation of the SLMCPR during a postulated CRWE event. Reference 27 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 23 is analyzed for the following operating conditions with applicable MCPR, APLHGR and LHGR limits.

x Base Case Operation x PROOS x SLO x Combined PROOS and TBVOOS x TBVOOS x Combined PROOS and FHOOS x FHOOS x Combined PROOS, TBVOOS and FHOOS x Combined TBVOOS and FHOOS Base Case Operation as well as the above-listed EOOS conditions assume all the items OOS below.

These conditions are general analysis assumptions used to ensure conservative analysis results and were not meant to define specific EOOS conditions beyond those already defined in Technical Specifications.

x Any 1 inoperable SRV x 2 inoperable TBV (Note that for TBVOOS, TBVOOS/FHOOS, PROOS/TBVOOS and PROOS/TBVOOS/FHOOS all 10 TBVs are assumed inoperable) x Up to 40% of the TIP channels OOS x Up to 50% of the LPRMs OOS Please note that during FFTR/Coastdown, FHOOS is included in Base Case Operation, TBVOOS, PROOS and PROOS/TBVOOS.

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 12, Revision 0 Single Loop Operation Brunswick Unit 2, Cycle 23 may operate in SLO up to a maximum core flow of 45 Mlbm/hr which corresponds to a maximum power level of 71.1% RTP with applicable MCPR, APLHGR and LHGR limits.

The following must be considered when operating in SLO:

x SLO is not permitted with RFWT (FHOOS).

x SLO is not permitted with TBVOOS.

x 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:

x The SLO Entry Rod Line is shown on the TLO maps to avoid regions of instability in the event of a pump trip.

x A maximum core flow line is shown on the SLO maps to avoid vibration problems.

x 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).

x 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 23 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:

x 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.

x Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10F and reactor power 23% RTP requires use of the TBVOOS/FHOOS limits.

x TBVOOS operation coincident with FHOOS is supported using the combined TBVOOS/FHOOS limits.

x TBVOOS operation coincident with PROOS is supported using the combined PROOS/TBVOOS limits.

x TBVOOS operation coincident with FHOOS and PROOS is supported using the combined PROOS/TBVOOS/FHOOS limits.

x SLO is not permitted with TBVOOS.

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 13, Revision 0 Feedwater Temperature Reduction Brunswick Unit 2, Cycle 23 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 - 10F. 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 110.3F. The following must be considered when operating with RFWT:

x 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.

x Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10F and reactor power 23% RTP requires use of the FHOOS limits.

x Until a core average exposure of EOCLB is reached, implementation of the FFTR/Coastdown limits is not required even if coastdown begins early.

x When operating with RFWT, the appropriate Stability Option III Power/Flow Maps (Figures 5 and

6) must be used.

x FHOOS operation coincident with TBVOOS is supported using the combined TBVOOS/FHOOS limits.

x FHOOS operation coincident with PROOS is supported using the combined PROOS/FHOOS limits.

x FHOOS operation coincident with TBVOOS and PROOS is supported using the combined PROOS/TBVOOS/FHOOS limits.

x SLO is not permitted with RFWT.

x NFWT limits have not been conservatively adjusted to eliminate the need to use RFWT limits below 50% RTP.

Pressure Regulator Out-of-Service Brunswick Unit 2, Cycle 23 may operate with one main turbine pressure regulator not available over the entire MEOD range and cycle with applicable APLHGR, MCPR and LHGR limits as specified in the COLR.

The following must be considered when operating with PROOS:

x Operation with the backup electro-hydraulic control main turbine pressure regulator not available requires the use of PROOS limits. The PROOS analysis supports operation with one pressure regulator not available.

x With TBVINS, prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10F and reactor power 23% RTP requires use of the PROOS/FHOOS limits.

x With TBVOOS, prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10F and reactor power 23% RTP requires use of the PROOS/TBVOOS/FHOOS limits.

x PROOS operation coincident with TBVOOS is supported using the combined PROOS/TBVOOS limits.

x PROOS operation coincident with FHOOS is supported using the combined PROOS/FHOOS limits.

x PROOS operation coincident with TBVOOS and FHOOS is supported using the combined PROOS/TBVOOS/FHOOS limits.

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 14, Revision 0 References In accordance with Brunswick Unit 2 Technical Specification 5.6.5.b, the analytical methods for determining Brunswick Unit 2 core operating limits have been specifically reviewed and approved by the NRC and are listed as References 1 through 21.

1. NEDE-24011-P-A, "GESTAR II - General Electric Standard Application for Reactor Fuel," and US Supplement, Revision 15, September 2005.

2. XN-NF-81-58(P)(A) and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Revision 2, March 1984.

3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Revision 1, September 1986.

4. 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.

5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs, Revision 1, May 1995.

6. XN-NF-80-19(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, March 1983.

7. 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.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Revision 0, October 1999.

9. 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.

10. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, February 1987.

11. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, June 2011.

12. ANF-913(P)(A) Volume 1 and Volume 1 Supplements 2, 3, 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Revision 1, August 1990.

13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors, Revision 3, September 2005.

14. EMF-2209(P)(A), SPCB Critical Power Correlation, Revision 3, September 2009.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Revision 0, August 2000.

16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model, Revision 0, May 2001.

17. EMF-2292(P)(A), ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients, Revision 0, September 2000.

18. EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2, Revision 0, August 2000.

19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, August 1996.

20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.

21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.

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Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 15, Revision 0

22. NEDC-31654P, Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, February 1989.

23. NEDO-31960-A, BWR Owners Group Long-Term Stability Solutions Licensing Methodology (Supplement 1), November 1995, NRC ADAMS Accession No. ML14093A211.

24. GENE-C51-00251-00-01, Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick 1 and 2, Revision 0, March 2001.

25. OG02-0119-260 Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy, July 17, 2002.

26. BAW-10255PA, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, Revision 2, May 2008.

27. BNP 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.

28. BNP Design Calculation 0B21-1015, BNP Power/Flow Maps, Revision 7, March 2008.

29. ANP-3560P, Brunswick Unit 2 Cycle 23 Reload Safety Analysis, Revision 0, January 2017.

30. BNP Design Calculation 2B21-1325, Preparation of the B2C23 Core Operating Limits Report, Revision 0.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 16 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 16, Revision 0 Table 1 RBM System Setpoints1 a Setpoint Value Allowable Value Setpoint b < 27.7 < 29.0 Lower Power Setpoint (LPSP )

b < 62.7 < 64.0 Intermediate Power Setpoint (IPSP )

b < 82.7 < 84.0 High Power Setpoint (HPSP )

c,d < 117.1 < 117.6 Low Trip Setpoint (LTSP )

c,d < 112.3 < 112.8 Intermediate Trip Setpoint (ITSP )

c,d < 107.3 < 107.8 High Trip Setpoint (HTSP )

RBM Time Delay (t d2 ) 0 seconds < 2.0 seconds a See Table 2 for RBM Operability Requirements.

b Setpoints in percent of Rated Thermal Power.

c Setpoints relative to a full scale reading of 125. For example, < 117.1 means

< 117.1/125.0 of full scale.

d Trip setpoints and allowable values are based on a HTSP Analytical Limit of 110.2 with RBM filter.

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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 17 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 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 1.86 TLO 29% and < 90%

1.89 SLO 90% 1.46 TLO 2

Requirements valid for all fuel designs, all SCRAM insertion times and all core average exposure ranges.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 18 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 18, Revision 0 Table 3 PBDA Setpoints3 Amplitude Trip OLMCPR(SS) OLMCPR(2PT)

Setpoint (Sp) 1.05 1.16 1.18 1.06 1.18 1.20 1.07 1.19 1.22 1.08 1.21 1.24 1.09 1.23 1.26 1.10 1.25 1.28 1.11 1.27 1.30 1.12 1.29 1.32 1.13 1.31 1.34 1.14 1.33 1.36 1.15 1.35 1.38 Acceptance Criteria Off-rated OLMCPR @ Rated Power 45% Flow OLMCPR PDBA Setpoint Setpoint Value Amplitude Trip (S p ) 1.10 Confirmation Count (N p ) 13 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 19 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 19, Revision 0 Table 4 Exposure Basis 4 for Brunswick Unit 2 Cycle 23 Transient Analysis Core Average Exposure Comments (MWd/MTU)

Breakpoint for design basis rod patterns to 35,915 5 EOFP + 15 EFPD (NEOC/EOCLB )

End of cycle with FFTR/Coastdown -

37,347 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.

5 NEOC exposure for Unit 2 Cycle 23 is defined as the same as the EOCLB exposure.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 20 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 20, Revision 0 Table 5 Power-Dependent MCPR p Limits 6 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPR p MCPR p 100.0 1.34 1.45 80.0 1.41 1.47 50.0 1.62 1.60 Base

> 65%F 65%F > 65%F 65%F Case 50.0 1.81 1.70 2.02 1.92 Operation 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.47 80.0 1.41 1.51 50.0 1.62 1.61

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.34 1.45 80.0 1.41 1.47 50.0 1.62 1.60

> 65%F 65%F > 65%F 65%F FHOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.34 1.45 80.0 1.42 1.53 80.0 1.55 1.70 50.0 1.81 2.02 PROOS > 65%F 65%F > 65%F 65%F 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 (continued) 6 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 21 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 21, Revision 0 Table 5 (continued)

Power-Dependent MCPR p Limits 7 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPR p MCPR p 100.0 1.37 1.47 80.0 1.41 1.51 TBVOOS 50.0 1.62 1.64 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 100.0 1.34 1.45 80.0 1.42 1.53 80.0 1.55 1.70 PROOS 50.0 1.81 2.02 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.47 80.0 1.42 1.53 80.0 1.55 1.70 PROOS 50.0 1.81 2.02 and > 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.37 1.47 80.0 1.42 1.53 PROOS 80.0 1.55 1.70 and 50.0 1.81 2.02 TBVOOS > 65%F 65%F > 65%F 65%F and 50.0 1.81 1.70 2.02 1.92 FHOOS 26.0 2.22 2.09 2.42 2.33 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 (end Table 5) 7 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 22 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 22, Revision 0 Table 6 Power-Dependent MCPR p Limits 8 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPR p MCPR p 100.0 1.37 1.48 80.0 1.41 1.49 50.0 1.62 1.60 Base

> 65%F 65%F > 65%F 65%F Case 50.0 1.82 1.72 2.03 1.93 Operation 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.39 1.50 80.0 1.43 1.54 50.0 1.62 1.63

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.37 1.48 80.0 1.41 1.49 50.0 1.62 1.60

> 65%F 65%F > 65%F 65%F FHOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.37 1.48 80.0 1.43 1.55 80.0 1.57 1.71 50.0 1.82 2.03 PROOS > 65%F 65%F > 65%F 65%F 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 (continued) 8 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 23 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 23, Revision 0 Table 6 (continued)

Power-Dependent MCPR p Limits 9 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPR p MCPR p 100.0 1.39 1.50 80.0 1.43 1.54 TBVOOS 50.0 1.62 1.66 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 100.0 1.37 1.48 80.0 1.43 1.55 80.0 1.57 1.71 PROOS 50.0 1.82 2.03 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 100.0 1.39 1.50 80.0 1.43 1.55 80.0 1.57 1.71 PROOS 50.0 1.82 2.03 and > 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 26.0 2.75 2.56 3.04 2.94 23.0 2.91 2.76 3.20 3.16 100.0 1.39 1.50 80.0 1.43 1.55 PROOS 80.0 1.57 1.71 and 50.0 1.82 2.03 TBVOOS > 65%F 65%F > 65%F 65%F and 50.0 1.82 1.72 2.03 1.93 FHOOS 26.0 2.23 2.12 2.43 2.35 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 (end Table 6) 9 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 24 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 24, Revision 0 Table 7 Power-Dependent MCPR p Limits 10 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPR p MCPR p Base Case 100.0 1.35 1.45 Operation 80.0 1.41 1.47 50.0 1.62 1.60 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 (Bounds operation 26.0 2.24 2.13 2.44 2.36 with NFWT) 23.0 2.33 2.21 2.50 2.44 TBVOOS 100.0 1.37 1.47 80.0 1.41 1.51 (FFTR/FHOOS 50.0 1.62 1.64 included) > 65%F 65%F > 65%F 65%F 50.0 1.81 1.70 2.02 1.92 (Bounds operation 26.0 2.22 2.09 2.42 2.33 with NFWT) 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 PROOS 100.0 1.35 1.45 80.0 1.42 1.53 (FFTR/FHOOS 80.0 1.55 1.70 included) 50.0 1.81 2.02

> 65%F 65%F > 65%F 65%F (Bounds operation 50.0 1.81 1.70 2.02 1.92 with NFWT) 26.0 2.22 2.09 2.42 2.33 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 PROOS 100.0 1.37 1.47 and 80.0 1.42 1.53 TBVOOS 80.0 1.55 1.70 50.0 1.81 2.02 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 1.81 1.70 2.02 1.92 26.0 2.22 2.09 2.42 2.33 (Bounds operation 26.0 2.82 2.69 3.18 3.10 with NFWT) 23.0 2.99 2.86 3.33 3.30 10 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 25 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 25, Revision 0 Table 8 Power-Dependent MCPR p Limits 11 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPR p MCPR p Base Case 100.0 1.37 1.48 Operation 80.0 1.41 1.49 50.0 1.62 1.60 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 (Bounds operation 26.0 2.24 2.13 2.44 2.36 with NFWT) 23.0 2.33 2.21 2.50 2.44 TBVOOS 100.0 1.39 1.50 80.0 1.43 1.54 (FFTR/FHOOS 50.0 1.62 1.66 included) > 65%F 65%F > 65%F 65%F 50.0 1.82 1.72 2.03 1.93 (Bounds operation 26.0 2.23 2.12 2.43 2.35 with NFWT) 26.0 2.82 2.69 3.18 3.10 23.0 2.99 2.86 3.33 3.30 PROOS 100.0 1.37 1.48 80.0 1.43 1.55 (FFTR/FHOOS 80.0 1.57 1.71 included) 50.0 1.82 2.03

> 65%F 65%F > 65%F 65%F (Bounds operation 50.0 1.82 1.72 2.03 1.93 with NFWT) 26.0 2.23 2.12 2.43 2.35 26.0 2.24 2.13 2.44 2.36 23.0 2.33 2.21 2.50 2.44 PROOS 100.0 1.39 1.50 and 80.0 1.43 1.55 TBVOOS 80.0 1.57 1.71 50.0 1.82 2.03 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 1.82 1.72 2.03 1.93 26.0 2.23 2.12 2.43 2.35 (Bounds operation 26.0 2.82 2.69 3.18 3.10 with NFWT) 23.0 2.99 2.86 3.33 3.30 11 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, 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.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 26 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 26, Revision 0 Table 9 Flow-Dependent MCPR f Limits 12 ATRIUM 10XM ATRIUM 11 LTAs Core Flow 13

(% of rated) MCPR f MCPR f 0.0 1.70 1.80 31.0 1.70 1.80 55.0 1.59 --

100.0 1.20 1.20 107.0 1.20 1.20 12 Limits valid for all SCRAM insertion times and all core average exposure ranges.

13

-- indicates that this fuel type does not have a breakpoint at the indicated exposure BNEI-0400-0005 Revision 0 l Attachment 1 l Page 27 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 27, Revision 0 Table 10 AREVA Fuel Steady-State LHGR SS Limits Peak ATRIUM 10XM ATRIUM 11 LTA 14 Pellet Exposure LHGR LHGR (GWd/MTU) (kW/ft) (kW/ft) 0.0 15.1 12.2 6.0 14.1 --

18.9 14.1 12.2 54.0 10.6 --

74.4 5.4 6.4 14

-- indicates that this fuel type does not have a breakpoint at the indicated exposure BNEI-0400-0005 Revision 0 l Attachment 1 l Page 28 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 28, Revision 0 Table 11 AREVA Fuel Power-Dependent LHGRFAC p Multipliers15 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base > 65%F 65%F > 65%F 65%F Case Operation 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 (continued) 15 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 29 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 29, Revision 0 Table 11 (continued)

AREVA Fuel Power-Dependent LHGRFAC p Multipliers16 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 TBVOOS 50.0 0.92 0.92 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 100.0 1.00 1.00 90.0 1.00 1.00 PROOS 50.0 0.86 0.86 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 PROOS 50.0 0.86 0.86 and > 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 PROOS 90.0 1.00 1.00 and 50.0 0.86 0.86 TBVOOS > 65%F 65%F > 65%F 65%F and 50.0 0.86 0.86 0.86 0.86 FHOOS 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 (end Table 11) 16 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 30 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 30, Revision 0 Table 12 AREVA Fuel Power-Dependent LHGRFAC p Multipliers17 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base > 65%F 65%F > 65%F 65%F Case Operation 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92

> 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F PROOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 (continued) 17 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 31 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 31, Revision 0 Table 12 (continued)

AREVA Fuel Power-Dependent LHGRFAC p Multipliers18 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp 100.0 1.00 1.00 90.0 1.00 1.00 TBVOOS 50.0 0.92 0.92 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 100.0 1.00 1.00 90.0 1.00 1.00 PROOS 50.0 0.86 0.86 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 PROOS 50.0 0.86 0.86 and > 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.43 0.50 0.43 0.50 23.0 0.40 0.46 0.40 0.46 100.0 1.00 1.00 PROOS 90.0 1.00 1.00 and 50.0 0.86 0.86 TBVOOS > 65%F 65%F > 65%F 65%F and 50.0 0.86 0.86 0.86 0.86 FHOOS 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 (end Table 12) 18 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 32 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 32, Revision 0 Table 13 AREVA Fuel Power-Dependent LHGRFAC p Multipliers19 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp Base Case 100.0 1.00 1.00 Operation 90.0 1.00 1.00 50.0 0.92 0.92 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 23.0 0.60 0.64 0.60 0.64 TBVOOS 100.0 1.00 1.00 90.0 1.00 1.00 (FFTR/FHOOS 50.0 0.92 0.92 included) > 65%F 65%F > 65%F 65%F 50.0 0.86 0.86 0.86 0.86 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 PROOS 100.0 1.00 1.00 90.0 1.00 1.00 (FFTR/FHOOS 50.0 0.86 0.86 included) > 65%F 65%F > 65%F 65%F 50.0 0.86 0.86 0.86 0.86 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 PROOS 100.0 1.00 1.00 and 90.0 1.00 1.00 TBVOOS 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F (FFTR/FHOOS 50.0 0.86 0.86 0.86 0.86 included) 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 (Bounds operation 23.0 0.38 0.43 0.38 0.43 with NFWT) 19 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 33 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 33, Revision 0 Table 14 AREVA Fuel Power-Dependent LHGRFAC p Multipliers 20 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) LHGRFACp LHGRFACp Base Case 100.0 1.00 1.00 Operation 90.0 1.00 1.00 50.0 0.92 0.92 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 23.0 0.60 0.64 0.60 0.64 TBVOOS 100.0 1.00 1.00 90.0 1.00 1.00 (FFTR/FHOOS 50.0 0.92 0.92 included) > 65%F 65%F > 65%F 65%F 50.0 0.86 0.86 0.86 0.86 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 26.0 0.40 0.46 0.40 0.46 23.0 0.38 0.43 0.38 0.43 PROOS 100.0 1.00 1.00 90.0 1.00 1.00 (FFTR/FHOOS 50.0 0.86 0.86 included) > 65%F 65%F > 65%F 65%F 50.0 0.86 0.86 0.86 0.86 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 PROOS 100.0 1.00 1.00 and 90.0 1.00 1.00 TBVOOS 50.0 0.86 0.86

> 65%F 65%F > 65%F 65%F (FFTR/FHOOS 50.0 0.86 0.86 0.86 0.86 included) 26.0 0.64 0.66 0.64 0.66 26.0 0.40 0.46 0.40 0.46 (Bounds operation 23.0 0.38 0.43 0.38 0.43 with NFWT) 20 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 34 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 34, Revision 0 Table 15 AREVA Fuel Flow-Dependent LHGRFAC f Multipliers 21 ATRIUM 10XM and Core Flow ATRIUM 11 LTA

(% of rated) LHGRFAC f 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 21 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 35 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 35, Revision 0 Table 16 AREVA Fuel Steady-State MAPLHGRSS Limits 22, 23 Average Planar ATRIUM 10XM ATRIUM 11 LTA Exposure MAPLHGR MAPLHGR (GWd/MTU) (kW/ft) (kW/ft) 0.0 13.1 10.5 15.0 13.1 10.5 67.0 7.7 5.9 22 AREVA Fuel MAPLHGR limits do not have a power or flow dependency.

23 ATRIUM 10XM and ATRIUM 11 MAPLHGR limits must be adjusted by a 0.80 multiplier when in SLO. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.

BNEI-0400-0005 Revision 0 l Attachment 1 l Page 36 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 36, Revision 0 Figure 1 Stability Option III Power/Flow Map OPRM Operable, Two Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0

 

   

APRM STP Scram  

110.0 

 

APRM STP Rod Block 100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 90.0 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 60.58 80.60 85 59.50 80.69 84 58.43 58 43 80.79 80 79

% Powerr 83 57.37 80.90 82 56.31 81.05 60.0 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 R 78 52.12 81.67 50.0 77 51.08 81.82 MELLL Line I e 76 50.05 81.98 SLO Entry Rod Line C g 75 49.02 82.13 74 48.00 82.29 40.0 F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Scram Avoidance Region n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 20.0 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 37 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 37, Revision 0 Figure 2 Stability Option III Power/Flow Map OPRM Inoperable, Two Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0

 

   

APRM STP Scram  

110.0 

 

APRM STP Rod Block 100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 90.0 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 60.58 80.60 85 59.50 80.69 84 58.43 58 43 80.79 80 79

% Power 83 57.37 80.90 60.0 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 R 78 52.12 81.67 50.0 I e 77 51.08 81.82 MELLL Line 76 50.05 81.98 SLO Entry Rod Line C g 75 49.02 82.13 74 48.00 82.29 40.0 Region A - Manual SCRAM F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Region B - Immediate Exit n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 5% Buffer Region 66 39.96 83.52 20.0 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 38 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 38, Revision 0 Figure 3 Stability Option III Power/Flow Map OPRM Operable, Single Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0

 

APRM STP Scram    

 

110.0 

 

100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 APRM 97 72.75 80.47 96 71.61 80.47 STP Rod 95 70.49 80.47 90.0 Block 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 60.58 80.60 85 59.50 80.69 84 58 43 58.43 80 79 80.79

% Power 83 57.37 80.90 82 56.31 81.05 60.0 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 R 78 52.12 81.67 50.0 77 51.08 81.82 MELLL Line I e 76 50.05 81.98 C g 75 49.02 82.13 74 48.00 82.29 40.0 F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Scram Avoidance Region n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 20.0 45 Mlb/hr Max Core Flow 65 38.97 83.68 64 37.99 83.83 Natural 63 37.01 83.99 Natural 62 36.04 84.14 Circulation OPRM Enabled Region 61 35.06 84.30 10.0 Circulation 60 34.10 84.45 Line Line 59 33.13 84.61 35% Minimum 35% Minimum Pump Pump Speed Speed Minimum Power Minimum Line Power Line 58 32.17 84.70 0.0 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 39 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 39, Revision 0 Figure 4 Stability Option III Power/Flow Map OPRM Inoperable, Single Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0

 

APRM STP Scram    

 

110.0 

 

100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 APRM 97 72.75 80.47 96 71.61 80.47 STP Rod 95 70.49 80.47 90.0 Block 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 60.58 80.60 85 59.50 80.69 84 58 43 58.43 80 79 80.79

% Power 83 57.37 80.90 60.0 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 R 78 52.12 81.67 50.0 77 51.08 81.82 MELLL Line I e 76 50.05 81.98 C g 75 49.02 82.13 74 48.00 82.29 40.0 Region A - Manual SCRAM F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Region B - Immediate Exit n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 5% Buffer Region 67 40.95 83.37 66 39.96 83.52 20.0 45 Mlb/hr Max Core Flow 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 40 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 40, Revision 0 Figure 5 Stability Option III Power/Flow Map OPRM Operable, FWTR, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0

 

   

APRM STP Scram  

110.0 

 

APRM STP Rod Block 100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 90.0 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 60.58 80.60 85 59.50 80.69 84 58.43 58 43 80.79 80 79

% Power 83 57.37 80.90 60.0 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 R 78 52.12 81.67 50.0 77 51.08 81.82 MELLL Line I e 76 50.05 81.98 SLO Entry Rod Line C g 75 49.02 82.13 (SLO prohibited 74 48.00 82.29 40.0 F i 73 46.98 82.44 during FWTR) 72 45.96 82.60 o

71 44.95 82.75 Scram Avoidance Region n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 20.0 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 41 of 42

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005 Page 41, Revision 0 Figure 6 Stability Option III Power/Flow Map OPRM Inoperable, FWTR, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0

 

   

APRM STP Scram  

110.0 

 

APRM STP Rod Block 100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 90.0 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 60.58 80.60 85 59.50 80.69 84 58.43 58 43 80.79 80 79

% Power 83 57.37 80.90 60.0 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 R 78 52.12 81.67 50.0 77 51.08 81.82 MELLL Line I e 76 50.05 81.98 SLO Entry Rod Line C g 75 49.02 82.13 (SLO prohibited 74 48.00 82.29 40.0 Region A - Manual SCRAM F i 73 46.98 82.44 during FWTR) 72 45.96 82.60 o

71 44.95 82.75 Region B - Immediate Exit n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 5% Buffer Region 66 39.96 83.52 20.0 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 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 BNEI-0400-0005 Revision 0 l Attachment 1 l Page 42 of 42