Text

Unit 1 Cycle 22 and Unit 2 Cycle 23 Core Operating Limits Reports
	ML18276A070
Person / Time
Site:	Brunswick
Issue date:	10/02/2018
From:	Wooten B; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA-18-0170
	Download: ML18276A070 (84)
	v • d • e

Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 October 2, 2018 Serial: RA-18-0170 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Unit 1 Cycle 22 and Unit 2 Cycle 23 Core Operating Limits Reports

References:

1. Letter from the U.S. Nuclear Regulatory Commission to Mr. William R.

Gideon, Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Regarding Core Flow Operating Range Expansion (MELLLA+),

dated September 18, 2018, ADAMS Accession Number ML18172A258

2. Letter from Bryan B. Wooten (Duke Energy) to the U.S. Nuclear Regulatory Commission Document Control Desk, Unit 1 Cycle 22 Core Operating Limits Report (COLR), dated March 27, 2018, ADAMS Accession Number ML18086B639

3. Letter from Mark McPherson (Duke Energy) to the U.S. Nuclear Regulatory Commission Document Control Desk, Unit 2 Cycle 23 Core Operating Limits Report (COLR), dated April 9, 2017, ADAMS Accession Number ML17100A840 Ladies and Gentlemen:

Enclosed are copies of the revised Core Operating Limits Reports (COLRs) for Brunswick Steam Electric Plant (BSEP) Unit 1 Cycle 22 and Unit 2 Cycle 23 operation. Duke Energy Progress, LLC (Duke Energy), is providing the enclosed COLRs in accordance with BSEP Unit 1 Technical Specification (TS) 5.6.5.d and Unit 2 TS 5.6.5.d. The enclosed COLRs were revised to support operation in the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain (i.e., Reference 1) and supersede in entirety the reports previously submitted by Reference 2 and 3.

No new regulatory commitments are contained in this letter.

RA-18-0170 Enclosure 1 Brunswick Unit 1 Cycle 22 Core Operating Limits Report, Dated September 2018

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 2 LIST OF EFFECTIVE PAGES Page(s) Revision 1- 39 1 This document consists of 39 total pages.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 3 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 CDA Setpoints................................................................................................................................ 10 RBM Setpoints ............................................................................................................................... 11 Equipment Out-of-Service .............................................................................................................. 11 Single Loop Operation.................................................................................................................... 12 Inoperable Main Turbine Bypass System ....................................................................................... 12 Feedwater Temperature Reduction ................................................................................................ 13 MELLLA+ Implementation .............................................................................................................. 13 References..................................................................................................................................... 14

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 4 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 3.1: BSP Endpoints for Nominal Feedwater Temperature .................................................... 18 Table 3.2: BSP Endpoints for Reduced Feedwater Temperature................................................... 18 Table 3.3: ABSP Setpoints for the Scram Region ......................................................................... 18 Table 4: Exposure Basis for Brunswick Unit 1 Cycle 22 Transient Analysis ................................ 19 Table 5: Power-Dependent MCPRp Limits .................................................................................. 20 NSS Insertion Times - BOC to < NEOC Table 6: Power-Dependent MCPRp Limits .................................................................................. 21 TSSS Insertion Times - BOC to < NEOC Table 7: Power-Dependent MCPRp Limits .................................................................................. 22 NSS Insertion Times - BOC to < EOCLB Table 8: Power-Dependent MCPRp Limits .................................................................................. 23 TSSS Insertion Times - BOC to < EOCLB Table 9: Power-Dependent MCPRp Limits .................................................................................. 24 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 10: Power-Dependent MCPRp Limits .................................................................................. 25 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 11: Flow-Dependent MCPRf Limits ..................................................................................... 26 Table 12: Framatome Fuel Steady-State LHGRSS Limits .............................................................. 27 Table 13: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 28 NSS Insertion Times - BOC to < EOCLB Table 14: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 29 TSSS Insertion Times - BOC to < EOCLB Table 15: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 30 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 16: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 31 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

Table 17: Framatome Fuel Flow-Dependent LHGRFACf Multipliers ............................................. 32 Table 18: Framatome Fuel Steady-State MAPLHGRSS Limits ...................................................... 33

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 5 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 DSS-CD Power/Flow Map ............................................................................... 34 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability DSS-CD Power/Flow Map ............................................................................... 35 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability DSS-CD Power/Flow Map ............................................................................... 36 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability DSS-CD Power/Flow Map ............................................................................... 37 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability DSS-CD Power/Flow Map ............................................................................... 38 OPRM Operable, FWTR, 2923 MWt Figure 6: Stability DSS-CD Power/Flow Map ............................................................................... 39 OPRM Inoperable, FWTR, 2923 MWt

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 6 NOMENCLATURE 2PT Two Recirculation Pump Trip W SLO Flow Uncertainty ABSP Automated Backup Stability Protection 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 CDA Confirmation Density Algorithm COLR Core Operating Limits Report CRWE Control Rod Withdrawal Error DSS-CD Detect and Suppress Solution - Confirmation Density ECCS Emergency Core Cooling System 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 HFCL High Flow Control Line 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 LHGRSS Steady-State Maximum Linear Heat Generation Rate LHGRFAC Linear Heat Generation Rate Factor LHGRFACf Flow-Dependent Linear Heat Generation Rate Factor LHGRFACp Power-Dependent Linear Heat Generation Rate Factor LOCA Loss of Coolant Accident LPRM Local Power Range Monitor (Subsystem)

LPSP Low Power Set Point LTSP Low Trip Set Point

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 7 NOMENCLATURE (continued)

MAPLHGR Maximum Average Planar Linear Heat Generation Rate MAPLHGRSS Steady-State Maximum Average Planar Linear Heat Generation Rate MAPFAC Maximum Average Planar Linear Heat Generation Rate Factor MAPFACf Flow-Dependent Maximum Average Planar Linear Heat Generation Rate Factor MAPFACp Power-Dependent Maximum Average Planar Linear Heat Generation Rate Factor MAPFACSLO Maximum Average Planar Linear Heat Generation Rate Factor when in SLO MCE Maximum Core Exposure MCPR Minimum Critical Power Ratio MCPRf Flow-Dependent Minimum Critical Power Ratio MCPRp Power-Dependent Minimum Critical Power Ratio MELLL Maximum Extended Load Line Limit MELLLA+ Maximum Extended Load Line Limit Analysis +

MEOD Maximum Extended Operating Domain MSIVOOS Main Steam Isolation Valve Out-of-Service NCL Natural Circulation Line 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)

PRNM Power Range Neutron Monitoring (System)

RBM Rod Block Monitor (Subsystem)

RDF Rated Drive Flow RFWT Reduced Feedwater Temperature RPT Recirculation Pump Trip RTP Rated Thermal Power SAD Amplitude Discriminator Setpoint (DSS-CD)

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

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 8 CAUTION References to COLR Figures or Tables should be made using titles only; Figure and Table numbers may change from cycle to cycle or between revisions.

Introduction and Summary The Brunswick Unit 1, Cycle 22 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, TS 3.2.2 LCO (MCPR) 9, 10, 11, TS 3.4.1 LCO (Recirculation loops 12, 13, 14, operating) 21 TS 3.7.6 LCO (Main Turbine bypass out-of-service)

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

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

4. The Manual Backup Stability 18, 19 TS Table 3.3.1.1-1, Function 2.f Protection (BSP) Scram Region (OPRM Upscale)

(Region I), Manual BSP Controlled Entry Region (Region II), the modified TS 3.3.1.1, Condition I and J (Alternate Average Power Range Monitor instability detection)

(APRM) Simulated Thermal Power -

High Scram setpoints used in the Automated BSP Scram Region, the BSP Boundary for TS 3.3.1.1.

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

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 B1C22, 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 DSS-CD stability ranges.

The generation of this COLR is documented in Reference 30 and is based on analysis results documented in References 23, 27-29.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 9 APLHGR Limits Steady-state MAPLHGRSS limits are provided for Framatome Fuel (Table 18). These steady-state MAPLHGRSS limits must be modified as follows:

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

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.

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

MAPLHGR LimitTLO = MAPLHGRSS MAPLHGR LimitSLO = MAPLHGRSS x MAPFACSLO where MAPFACSLO = 0.80 for ATRIUM 10XM 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 11 are based on the TLO and SLO SLMCPRs listed in Technical Specification 2.1.1.2 of 1.07 and 1.09, respectively.

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), 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.

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.

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 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.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 10 LHGR Limits Steady-state LHGRSS limits are provided for Framatome Fuel (Table 12). These steady-state LHGRSS limits must be modified as follows:

Framatome Fuel LHGR limits have a core power and core flow dependency. Framatome Fuel power-dependent LHGRFACp multipliers (Tables 13-16) and flow-dependent LHGRFACf multipliers (Table 17) must be used to modify the steady-state LHGRSS limits (Table 12) for off-rated conditions.

Framatome Fuel power-dependent LHGRFACp 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.

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 = LHGRSS x (LHGRFACp, LHGRFACf)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.

The cycle-specific off-rated flow dependent LHGR set-down bounds those assumed in the MELLLA+

plant-specific ECCS-LOCA analyses.

CDA Setpoints Brunswick Unit 1 has implemented the stability Detect and Suppress Solution - Confirmation Density (DSS-CD) solution using the Oscillation Power Range Monitor (OPRM) as described in Reference 19.

Plant-specific analyses for the DSS-CD Solution are provided in Reference 23. The Detect and Suppress function of the DSS-CD solution based on the OPRM system relies on the Confirmation Density Algorithm (CDA), which constitutes the licensing basis. The Backup Stability Protection (BSP) solution may be used by the plant in the event the OPRM Upscale function is declared inoperable.

The CDA enabled through the OPRM system and the BSP solution described in Reference 23 provide the stability licensing bases for B1C22. The safety evaluation report for Reference 19 concluded that the DSS-CD solution is acceptable subject to certain cycle-specific limitations and conditions. These cycle-specific limitations and conditions are met for B1C22.

A reload DSS-CD evaluation has been performed in accordance with the licensing methodology described in Reference 19 to confirm the Amplitude Discriminator Setpoint (SAD) of the CDA established in Reference 23. The Cycle 22 DSS-CD evaluation demonstrates that: 1) the DSS-CD Solution is applicable to B1C22; and, 2) the SAD value of 1.10 established in Reference 23 is confirmed for operation of B1C22.

The SAD setpoint value of 1.10 is applicable to TLO and to SLO.

Reference 19 describes two BSP options that are based on selected elements from three distinct constituents: BSP Manual Regions, BSP Boundary, and Automated BSP (ABSP) setpoints.

The Manual BSP region boundaries and the BSP Boundary were calculated for Brunswick Unit 1 Cycle 22 for nominal feedwater temperature operation and reduced feedwater temperature. The endpoints of the regions are defined in Table 3.1 and Table 3.2. The Manual BSP region boundary endpoints are calculated with the Reference 18 methodology and connected using the Generic Shape Function (GSF),

which is described in Reference 19.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 11 The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are determined for Cycle 22 and are defined in Table 3.3. These ABSP setpoints are applicable to both TLO and SLO as well as nominal and reduced feedwater temperature operation.

The Manual Backup Stability Protection (BSP) Regions I and II are documented on the Power/Flow maps as is the modified APRM Simulated Thermal Power (STP) high SCRAM setpoints and the BSP Boundary.

The power/flow maps (Figures 1-6) were validated for B1C22 based on Reference 29 using the Reference 19 methodology to facilitate operation under DSS-CD as implemented by Function 2.f of Table 3.3.1.1-1 and LCO Conditions I and J 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 23% RTP and below 75% drive flow (correlated to core flow) where the OPRM 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 there is increased probability the OPRM system may generate a scram to avoid an instability event. Figures 2, 4, and 6 support an inoperable OPRM and highlight the Manual Backup Stability Regions I and II, the modified APRM STP high SCRAM setpoints, and the BSP Boundary. 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-217345 which 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 1, Cycle 22 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 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.

Any 1 inoperable SRV

1 inoperable TBV (Note that for TBVOOS, TBVOOS/FHOOS, all 4 TBVs are assumed inoperable)

Up to 40% of the TIP channels OOS

Up to 50% of the LPRMs OOS

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 12 Please note that during FFTR/Coastdown, FHOOS is included in Base Case Operation and TBVOOS.

Single Loop Operation Brunswick Unit 1, Cycle 22 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.

These power and flow limitations also apply when operating with jet pump loop flow mismatch conditions (LCO 3.4.1). The following must be considered when operating in SLO:

SLO is not permitted with RFWT (FHOOS/FFTR).

SLO is not permitted with TBVOOS.

SLO is not permitted with MSIVOOS.

SLO is not permitted within the MELLLA+ operating domain.

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).

When in SLO, Figures 3 and 4 implement the corrective action for AR-217345 which 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.

If OPRMs are inoperable in SLO, the expansion of the ABSP region results in power being restricted to 39% RTP as shown in Figure 4.

Inoperable Main Turbine Bypass System Brunswick Unit 1, Cycle 22 may operate with an inoperable Main Turbine Bypass System over the entire MEOD range and cycle and in the MELLLA+ domain after achieving 4.75 GWd/MTU cycle exposure with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. An operable Main Turbine Bypass System with only one inoperable bypass valve 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:

Two 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.

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

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

SLO is not permitted with TBVOOS.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 13 Feedwater Temperature Reduction Brunswick Unit 1, Cycle 22 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:

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 >10F and reactor power 23% RTP requires use of the 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 DSS-CD 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.

FWTR operation within the MELLLA+ operating domain is not allowed.

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

MELLLA+ Implementation Revision 1 of this COLR is being issued to support implementation of B1C22 operation in the MELLLA+

domain. Only MELLLA domain operation is allowed prior to a cycle exposure of 4.75 GWd/MTU. Either MELLLA or MELLLA+ domain operation is allowed after 4.75 GWd/MTU cycle exposure. This restriction is administratively controlled.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 14 References In accordance with Brunswick Unit 1 Technical Specification 5.6.5.b, the analytical methods for determining Brunswick Unit 1 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, Revision 0, 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. NEDC-33075P-A, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, Revision 8, November 2013.

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.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 15

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

23. Safety Analysis Report For Brunswick Steam Electric Plants Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus, DUKE-0B21-1104-000(P), July 2016.

24. Not Used.

25. Not Used.

26. Not Used.

27. BNP Design Calculation 1C51-0001, Power Range Neutron Monitoring System Setpoint Uncertainty and Scaling Calculation (1-C51-APRM-1 through 4 Loops and 1-C51 RBM-A and B Loops),

Revision 4, September 2018.

28. BNP Design Calculation 0B21-2045, BNP Power/Flow Maps For MELLLA+, Revision 1, September 2017.

29. ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0, December 2017.

30. BNP Design Calculation 1B21-2050, Preparation of the B1C22 Core Operating Limits Report, Revision 1, September 2018.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 16 Table 1 RBM System Setpoints1 Setpoint a Setpoint Value Allowable Value Lower Power Setpoint (LPSPb) < 27.7 < 29.0 Intermediate Power Setpoint (IPSPb) < 62.7 < 64.0 High Power Setpoint (HPSPb) < 82.7 < 84.0 Low Trip Setpoint (LTSPc,d) < 117.1 < 117.6 Intermediate Trip Setpoint (ITSPc,d) < 112.3 < 112.8 High Trip Setpoint (HTSPc,d) < 107.3 < 107.8 RBM Time Delay (td2) 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.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 17 Table 2 RBM Operability Requirements 2 IF the following conditions are met, THEN RBM Not Required Operable Thermal Power

(% rated) MCPR 1.89 TLO 29% and < 90%

1.92 SLO 90% 1.49 TLO 2 Requirements valid for all fuel designs, all SCRAM insertion times and all core average exposure ranges.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 18 Table 3.1 BSP Endpoints for Nominal Feedwater Temperature 3,4 Power Flow Endpoint Definition

(%) (%)

Scram Region A1 57.0 40.6 Boundary, HFCL Scram Region B1 42.0 31.7 Boundary, NCL Controlled Entry Region A2 64.5 50.0 Boundary, HFCL Controlled Entry Region B2 28.9 31.9 Boundary, NCL Table 3.2 BSP Endpoints for Reduced Feedwater Temperature3,4 Power Flow Endpoint Definition

(%) (%)

Scram Region A1 65.9 51.8 Boundary, HFCL Scram Region B1 36.5 31.9 Boundary, NCL Controlled Entry Region A2 69.8 56.8 Boundary, HFCL Controlled Entry Region B2 28.9 31.9 Boundary, NCL Table 3.3 ABSP Setpoints for the Scram Region3,5 Parameter Symbol Value Slope of ABSP APRM flow-biased trip linear segment.

mTRIP 2.00 ABSP APRM flow-biased trip setpoint power intercept. Constant PBSP-TRIP 42.0 %RTP Power Line for Trip from zero Drive Flow to Flow Breakpoint value.

ABSP APRM flow-biased trip WBSP-TRIP setpoint drive flow intercept. 37.5 %RDF Constant Flow Line for Trip.

Flow Breakpoint value WBSP-BREAK 25.0 %RDF 3 These tables are referred to by Technical Specification 3.3.1.1 (Table 3.3.1.1-1) and 5.6.5.a.4.

4 The BSP Boundary for Nominal and Reduced Feedwater Temperature is defined by the MELLLA boundary line and extends from the natural circulation boundary to rated power.

5 When in SLO the ABSP STP Scram is modified by the applied SLO W as shown in Figure 4.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 19 Table 4 Exposure Basis6 for Brunswick Unit 1 Cycle 22 Transient Analysis Core Average Exposure Comments (MWd/MTU)

Breakpoint for exposure dependent MCPRp 34,284 limits (NEOC)

Design basis rod patterns to EOFP + 15 35,484 EFPD (EOCLB)

End of cycle with FFTR/Coastdown -

36,948 Maximum Core Exposure (MCE) 6 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.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 20 Table 5 Power-Dependent MCPRp Limits 7 NSS Insertion Times BOC to < NEOC EOOS Power ATRIUM 10XM

(%

Condition MCPRp rated) 100.0 1.35 90.0 1.37 Base 50.0 1.66 Case > 65%F 65%F Operation 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.38 90.0 1.40 50.0 1.66 TBVOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.35 90.0 1.37 50.0 1.66 FHOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20 7 Limits support operation with any combination of any 1 inoperable SRV, 1 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.

FHOOS not permitted in the MELLLA+ domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 21 Table 6 Power-Dependent MCPRp Limits 8 TSSS Insertion Times BOC to < NEOC EOOS Power ATRIUM 10XM

(%

Condition MCPRp rated) 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F Case Operation 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.41 90.0 1.44 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.39 90.0 1.40 50.0 1.66

> 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20 8 Limits support operation with any combination of any 1 inoperable SRV, 1 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.

FHOOS not permitted in the MELLLA+ domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 22 Table 7 Power-Dependent MCPRp Limits 9 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM

(%

Condition MCPRp rated) 100.0 1.36 90.0 1.37 Base 50.0 1.66 Case > 65%F 65%F Operation 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.38 90.0 1.40 50.0 1.66 TBVOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.36 90.0 1.37 50.0 1.66 FHOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20 9 Limits support operation with any combination of any 1 inoperable SRV, 1 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.

FHOOS not permitted in the MELLLA+ domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 23 Table 8 Power-Dependent MCPRp Limits 10 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM

(%

Condition MCPRp rated) 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F Case Operation 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.41 90.0 1.44 50.0 1.66

> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.39 90.0 1.40 50.0 1.66

> 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20 10 Limits support operation with any combination of any 1 inoperable SRV, 1 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.

FHOOS not permitted in the MELLLA+ domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 24 Table 9 Power-Dependent MCPRp Limits 11 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM

(%

Condition MCPRp rated)

Base Case 100.0 1.37 Operation 90.0 1.37 50.0 1.66 (FFTR/FHOOS > 65%F 65%F included) 50.0 1.91 1.78 26.0 2.34 2.22 (Bounds 26.0 2.51 2.46 operation with NFWT) 23.0 2.60 2.59 TBVOOS 100.0 1.38 90.0 1.40 (FFTR/FHOOS 50.0 1.66 included) > 65%F 65%F 50.0 1.91 1.78 (Bounds 26.0 2.34 2.22 operation with NFWT) 26.0 3.03 2.96 23.0 3.22 3.20 11 Limits support operation with any combination of any 1 inoperable SRV, 1 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.

FHOOS not permitted in the MELLLA+ domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 25 Table 10 Power-Dependent MCPRp Limits 12 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM

(%

Condition MCPRp rated)

Base Case 100.0 1.39 Operation 90.0 1.40 50.0 1.66 (FFTR/FHOOS > 65%F 65%F included) 50.0 1.93 1.80 26.0 2.36 2.25 (Bounds 26.0 2.51 2.46 operation with NFWT) 23.0 2.60 2.59 TBVOOS 100.0 1.41 90.0 1.44 (FFTR/FHOOS 50.0 1.66 included) > 65%F 65%F 50.0 1.93 1.80 (Bounds 26.0 2.36 2.25 operation with NFWT) 26.0 3.03 2.96 23.0 3.22 3.20 12 Limits support operation with any combination of any 1 inoperable SRV, 1 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.

FHOOS not permitted in the MELLLA+ domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 26 Table 11 Flow-Dependent MCPRf Limits 13 Core Flow ATRIUM 10XM

(% of rated) MCPRf 0.0 1.64 31.0 1.64 60.0 1.50 80.0 1.30 100.0 1.30 107.0 1.30 13 Limits valid for all SCRAM insertion times, all core average exposure ranges, all EOOS scenarios, and both TLO

& SLO.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 27 Table 12 Framatome Fuel Steady-State LHGRSS Limits Peak ATRIUM 10XM Pellet Exposure LHGR (GWd/MTU) (kW/ft) 0.0 14.1 6.0 14.1 18.9 14.1 54.0 10.6 74.4 5.4

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 28 Table 13 Framatome Fuel Power-Dependent LHGRFACp Multipliers 14 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM

(%

Condition LHGRFACp rated) 100.0 1.00 90.0 1.00 50.0 1.00 Base > 65%F 65%F Case Operation 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.51 0.53 23.0 0.49 0.50 100.0 1.00 90.0 1.00 50.0 1.00

> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.42 0.48 23.0 0.39 0.43 100.0 1.00 90.0 1.00 50.0 0.97

> 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 TBVOOS 50.0 0.96 and > 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40 14 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. FHOOS not permitted in the MELLLA+

domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 29 Table 14 Framatome Fuel Power-Dependent LHGRFACp Multipliers 15 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM

(%

Condition LHGRFACp rated) 100.0 1.00 90.0 1.00 50.0 1.00 Base > 65%F 65%F Case Operation 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.51 0.53 23.0 0.49 0.50 100.0 1.00 90.0 1.00 50.0 1.00

> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.42 0.48 23.0 0.39 0.43 100.0 1.00 90.0 1.00 50.0 0.97

> 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 TBVOOS 50.0 0.96 and > 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40 15 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. FHOOS not permitted in the MELLLA+

domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 30 Table 15 Framatome Fuel Power-Dependent LHGRFACp Multipliers 16 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM

(%

Condition LHGRFACp rated)

Base Case 100.0 1.00 Operation 90.0 1.00 50.0 0.97 (FFTR/FHOOS > 65%F 65%F included) 50.0 0.89 0.95 26.0 0.63 0.77 (Bounds 26.0 0.46 0.48 operation with NFWT) 23.0 0.44 0.46 TBVOOS 100.0 1.00 90.0 1.00 (FFTR/FHOOS 50.0 0.96 included) > 65%F 65%F 50.0 0.89 0.95 (Bounds 26.0 0.63 0.77 operation with NFWT) 26.0 0.39 0.44 23.0 0.37 0.40 16 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. FHOOS not permitted in the MELLLA+

domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 31 Table 16 Framatome Fuel Power-Dependent LHGRFACp Multipliers 17 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)

EOOS Power ATRIUM 10XM

(%

Condition LHGRFACp rated)

Base Case 100.0 1.00 Operation 90.0 1.00 50.0 0.97 (FFTR/FHOOS > 65%F 65%F included) 50.0 0.89 0.95 26.0 0.63 0.77 (Bounds 26.0 0.46 0.48 operation with NFWT) 23.0 0.44 0.46 TBVOOS 100.0 1.00 90.0 1.00 (FFTR/FHOOS 50.0 0.96 included) > 65%F 65%F 50.0 0.89 0.95 (Bounds 26.0 0.63 0.77 operation with NFWT) 26.0 0.39 0.44 23.0 0.37 0.40 17 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. FHOOS not permitted in the MELLLA+

domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 32 Table 17 Framatome Fuel Flow-Dependent LHGRFACf Multipliers 18 Core Flow ATRIUM 10XM

(% of rated) LHGRFACf 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 18 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 33 Table 18 Framatome Fuel Steady-State MAPLHGRSS Limits19, 20 Average Planar ATRIUM 10XM Exposure MAPLHGR (GWd/MTU) (kW/ft) 0.0 13.1 15.0 13.1 67.0 7.7 19 Framatome Fuel MAPLHGR limits do not have a power, flow, or EOOS dependency.

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

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 34 Figure 1 Stability DSS-CD 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 Minimum Minimum Maximum (MELLLA) (MELLLA+) (ICF)

APRM STP Scram Core Core Core 110.0 MELLLA+ Line Power Flow Flow Flow

% Mlbs/hr Mlbs/hr Mlbs/hr APRM STP Rod Block 100 76.19 65.45 80.47 99 75.04 64.42 80.47 100.0 98 73.89 63.39 80.47 97 72.75 62.35 80.47 96 71.61 61.32 80.47 95 70.49 60.29 80.47 90.0 94 69.36 59.26 80.47 93 68.25 58.22 80.47 92 67.13 57.19 80.47 91 66.03 56.16 80.47 80.0 90 64.93 55.13 80.47 89 63.83 54.10 80.47 88 62.74 53.06 80.47 87 61.66 52.03 80.51 70.0 86 60.58 51.00 80.60 85 59.50 49.97 80.69 84 58.43 48.94 80.79

% Power 83 57.37 47.90 80.90 60.0 82 56.31 46.87 81.05 81 55.25 45.84 81.21 80 54.20 44.81 81.36 79 53.16 43.77 81.51 R 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary SLO Entry 76 50.05 --- 81.98 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 Scram Avoidance Region 71 44.95 --- 82.75 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 OPRM Enabled Region 62 36.04 --- 84.14 61 35.06 --- 84.30 10.0 Circulation 60 34.10 --- 84.45 Line 35% Approximate 59 33.13 --- 84.61 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:

0B21-2045, Revision 1

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 35 Figure 2 Stability DSS-CD 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 Minimum Minimum Maximum (MELLLA) (MELLLA+) (ICF)

APRM STP Scram Core Core Core 110.0 MELLLA+ Line Power Flow Flow Flow

% Mlbs/hr Mlbs/hr Mlbs/hr APRM STP Rod Block 100 76.19 65.45 80.47 99 75.04 64.42 80.47 100.0 98 73.89 63.39 80.47 97 72.75 62.35 80.47 96 71.61 61.32 80.47 95 70.49 60.29 80.47 90.0 94 69.36 59.26 80.47 93 68.25 58.22 80.47 92 67.13 57.19 80.47 ABSP STP Scram 91 66.03 56.16 80.47 80.0 90 64.93 55.13 80.47 89 63.83 54.10 80.47 ABSP STP Rod Block 88 62.74 53.06 80.47 87 61.66 52.03 80.51 70.0 86 60.58 51.00 80.60 85 59.50 49.97 80.69 84 58.43 48.94 80.79

% Power 83 57.37 47.90 80.90 82 56.31 46.87 81.05 60.0 81 55.25 45.84 81.21 80 54.20 44.81 81.36 79 53.16 43.77 81.51 R 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary SLO Entry 76 50.05 --- 81.98 Rod Line C g 75 49.02 --- 82.13 74 48.00 --- 82.29 40.0 Region I - Manual Scram F i 73 46.98 --- 82.44 o 72 45.96 --- 82.60 71 44.95 --- 82.75 Region II - Controlled Entry n 70 43.94 --- 82.91 30.0 69 42.94 --- 83.06 68 41.94 --- 83.22 Operator Awareness 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 OPRM Enabled Region 62 36.04 --- 84.14 61 35.06 --- 84.30 10.0 Circulation 60 34.10 --- 84.45 Line 35% Approximate 59 33.13 --- 84.61 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:

0B21-2045, Revision 1

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 36 Figure 3 Stability DSS-CD 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 Minimum Minimum Maximum MELLLA+ Line (MELLLA) (MELLLA+) (ICF)

(MELLLA+ Operations APRM STP Scram Core Core Core 110.0 prohibited during SLO) Power Flow Flow Flow

% Mlbs/hr Mlbs/hr Mlbs/hr 100 76.19 65.45 80.47 99 75.04 64.42 80.47 100.0 98 73.89 63.39 80.47 APRM 97 72.75 62.35 80.47 96 71.61 61.32 80.47 STP Rod 95 70.49 60.29 80.47 90.0 Block 94 69.36 59.26 80.47 93 68.25 58.22 80.47 92 67.13 57.19 80.47 91 66.03 56.16 80.47 80.0 90 64.93 55.13 80.47 89 63.83 54.10 80.47 88 62.74 53.06 80.47 87 61.66 52.03 80.51 70.0 86 60.58 51.00 80.60 85 59.50 49.97 80.69 84 58.43 48.94 80.79

% Power 83 57.37 47.90 80.90 82 56.31 46.87 81.05 60.0 81 55.25 45.84 81.21 80 54.20 44.81 81.36 79 53.16 43.77 81.51 R 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary 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 Scram Avoidance Region 71 44.95 --- 82.75 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 63 37.01 --- 83.99 Natural OPRM Enabled Region 62 36.04 --- 84.14 61 35.06 --- 84.30 10.0 Circulation 60 34.10 --- 84.45 Line 35% Approximate 59 33.13 --- 84.61 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:

0B21-2045, Revision 1

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 37 Figure 4 Stability DSS-CD 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 Minimum Minimum Maximum MELLLA+ Line (MELLLA) (MELLLA+) (ICF)

(MELLLA+ Operations APRM STP Scram Core Core Core 110.0 prohibited during SLO) Power Flow Flow Flow

% Mlbs/hr Mlbs/hr Mlbs/hr 100 76.19 65.45 80.47 99 75.04 64.42 80.47 100.0 98 73.89 63.39 80.47 APRM 97 72.75 62.35 80.47 96 71.61 61.32 80.47 STP Rod 95 70.49 60.29 80.47 90.0 Block 94 69.36 59.26 80.47 93 68.25 58.22 80.47 ABSP STP Scram 92 67.13 57.19 80.47 91 66.03 56.16 80.47 80.0 90 64.93 55.13 80.47 89 63.83 54.10 80.47 ABSP STP Rod Block 88 62.74 53.06 80.47 87 61.66 52.03 80.51 70.0 86 60.58 51.00 80.60 85 59.50 49.97 80.69 84 58.43 48.94 80.79

% Power 83 57.37 47.90 80.90 82 56.31 46.87 81.05 60.0 81 55.25 45.84 81.21 80 54.20 44.81 81.36 79 53.16 43.77 81.51 R 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary 76 50.05 --- 81.98 C g 75 49.02 --- 82.13 74 48.00 --- 82.29 40.0 Region I - Manual Scram F i 73 46.98 --- 82.44 o 72 45.96 --- 82.60 71 44.95 --- 82.75 Region II - Controlled Entry n 70 43.94 --- 82.91 30.0 69 42.94 --- 83.06 68 41.94 --- 83.22 Operator Awareness 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 OPRM Enabled Region 62 36.04 --- 84.14 61 35.06 --- 84.30 10.0 Circulation 60 34.10 --- 84.45 Line 35% Approximate 59 33.13 --- 84.61 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:

0B21-2045, Revision 1

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 38 Figure 5 Stability DSS-CD 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 Minimum Minimum Maximum MELLLA+ Line (MELLLA) (MELLLA+) (ICF)

(MELLLA+ Operations APRM STP Scram Core Core Core 110.0 prohibited during FWTR) Power Flow Flow Flow

% Mlbs/hr Mlbs/hr Mlbs/hr APRM STP Rod Block 100 76.19 65.45 80.47 99 75.04 64.42 80.47 100.0 98 73.89 63.39 80.47 97 72.75 62.35 80.47 96 71.61 61.32 80.47 95 70.49 60.29 80.47 90.0 94 69.36 59.26 80.47 93 68.25 58.22 80.47 92 67.13 57.19 80.47 91 66.03 56.16 80.47 80.0 90 64.93 55.13 80.47 89 63.83 54.10 80.47 88 62.74 53.06 80.47 87 61.66 52.03 80.51 70.0 86 60.58 51.00 80.60 85 59.50 49.97 80.69 84 58.43 48.94 80.79

% Power 83 57.37 47.90 80.90 60.0 82 56.31 46.87 81.05 81 55.25 45.84 81.21 80 54.20 44.81 81.36 79 53.16 43.77 81.51 R 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary SLO Entry 76 50.05 --- 81.98 Rod Line C g 75 49.02 --- 82.13 74 48.00 --- 82.29 40.0 (SLO prohibited F i 73 46.98 --- 82.44 during FWTR) o 72 45.96 --- 82.60 Scram Avoidance Region 71 44.95 --- 82.75 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 OPRM Enabled Region 62 36.04 --- 84.14 61 35.06 --- 84.30 10.0 Circulation 60 34.10 --- 84.45 Line 35% Approximate 59 33.13 --- 84.61 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:

0B21-2045, Revision 1

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 1 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 1 Page 39 Figure 6 Stability DSS-CD 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 Minimum Minimum Maximum MELLLA+ Line (MELLLA) (MELLLA+) (ICF)

(MELLLA+ Operations APRM STP Scram Core Core Core 110.0 prohibited during FWTR) Power Flow Flow Flow

% Mlbs/hr Mlbs/hr Mlbs/hr APRM STP Rod Block 100 76.19 65.45 80.47 99 75.04 64.42 80.47 100.0 98 73.89 63.39 80.47 97 72.75 62.35 80.47 96 71.61 61.32 80.47 95 70.49 60.29 80.47 90.0 94 69.36 59.26 80.47 93 68.25 58.22 80.47 92 67.13 57.19 80.47 ABSP STP Scram 91 66.03 56.16 80.47 80.0 90 64.93 55.13 80.47 89 63.83 54.10 80.47 ABSP STP Rod Block 88 62.74 53.06 80.47 87 61.66 52.03 80.51 70.0 86 60.58 51.00 80.60 85 59.50 49.97 80.69 84 58.43 48.94 80.79

% Power 83 57.37 47.90 80.90 82 56.31 46.87 81.05 60.0 81 55.25 45.84 81.21 80 54.20 44.81 81.36 79 53.16 43.77 81.51 R 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary SLO Entry 76 50.05 --- 81.98 Rod Line C g 75 49.02 --- 82.13 74 48.00 --- 82.29 40.0 Region I - Manual Scram (SLO prohibited F i 73 46.98 --- 82.44 during FWTR) o 72 45.96 --- 82.60 71 44.95 --- 82.75 Region II - Controlled Entry n 70 43.94 --- 82.91 30.0 69 42.94 --- 83.06 68 41.94 --- 83.22 Operator Awareness 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 OPRM Enabled Region 62 36.04 --- 84.14 61 35.06 --- 84.30 10.0 Circulation 60 34.10 --- 84.45 Line 35% Approximate 59 33.13 --- 84.61 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:

0B21-2045, Revision 1

RA-18-0170 Enclosure 2 Brunswick Unit 2 Cycle 23 Core Operating Limits Report, Dated September 2018

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 2, Revision 1 LIST OF EFFECTIVE PAGES Page(s) Revision 1- 41 1 This document consists of 41 total pages.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 3, Revision 1 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 CDA 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 MELLLA+ Implementation .............................................................................................................. 13 References..................................................................................................................................... 14

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 4, Revision 1 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 3.1: BSP Endpoints for Nominal Feedwater Temperature .................................................... 18 Table 3.2: BSP Endpoints for Reduced Feedwater Temperature................................................... 18 Table 3.3: ABSP Setpoints for the Scram Region ......................................................................... 18 Table 4: Exposure Basis for Brunswick Unit 2 Cycle 23 Transient Analysis ................................ 19 Table 5: Power-Dependent MCPRp Limits .................................................................................. 20 NSS Insertion Times - BOC to < EOCLB Table 6: Power-Dependent MCPRp Limits .................................................................................. 22 TSSS Insertion Times - BOC to < EOCLB Table 7: Power-Dependent MCPRp Limits .................................................................................. 24 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

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

Table 9: Flow-Dependent MCPRf Limits ..................................................................................... 26 Table 10: Framatome Fuel Steady-State LHGRSS Limits .............................................................. 27 Table 11: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 28 NSS Insertion Times - BOC to < EOCLB Table 12: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 30 TSSS Insertion Times - BOC to < EOCLB Table 13: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 32 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)

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

Table 15: Framatome Fuel Flow-Dependent LHGRFACf Multipliers ............................................. 34 Table 16: Framatome Fuel Steady-State MAPLHGRSS Limits ...................................................... 35

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 5, Revision 1 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 DSS-CD Power/Flow Map ............................................................................... 36 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability DSS-CD Power/Flow Map ............................................................................... 37 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability DSS-CD Power/Flow Map ............................................................................... 38 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability DSS-CD Power/Flow Map ............................................................................... 39 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability DSS-CD Power/Flow Map ............................................................................... 40 OPRM Operable, FWTR, 2923 MWt Figure 6: Stability DSS-CD Power/Flow Map ............................................................................... 41 OPRM Inoperable, FWTR, 2923 MWt

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 6, Revision 1 NOMENCLATURE 2PT Two Recirculation Pump Trip W SLO Flow Uncertainty ABSP Automated Backup Stability Protection 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 CDA Confirmation Density Algorithm COLR Core Operating Limits Report CRWE Control Rod Withdrawal Error DSS-CD Detect and Suppress Solution - Confirmation Density ECCS Emergency Core Cooling System 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 HFCL High Flow Control Line 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 LHGRSS Steady-State Maximum Linear Heat Generation Rate LHGRFAC Linear Heat Generation Rate Factor LHGRFACf Flow-Dependent Linear Heat Generation Rate Factor LHGRFACp Power-Dependent Linear Heat Generation Rate Factor LOCA Loss of Coolant Accident LPRM Local Power Range Monitor (Subsystem)

LPSP Low Power Set Point LTA Lead Test Assembly LTSP Low Trip Set Point

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 7, Revision 1 NOMENCLATURE (continued)

MAPLHGR Maximum Average Planar Linear Heat Generation Rate MAPLHGRSS Steady-State Maximum Average Planar Linear Heat Generation Rate MAPFAC Maximum Average Planar Linear Heat Generation Rate Factor MAPFACf Flow-Dependent Maximum Average Planar Linear Heat Generation Rate Factor MAPFACp Power-Dependent Maximum Average Planar Linear Heat Generation Rate Factor MAPFACSLO Maximum Average Planar Linear Heat Generation Rate Factor when in SLO MCE Maximum Core Exposure MCPR Minimum Critical Power Ratio MCPRf Flow-Dependent Minimum Critical Power Ratio MCPRp Power-Dependent Minimum Critical Power Ratio MELLL Maximum Extended Load Line Limit MELLLA+ Maximum Extended Load Line Limit Analysis +

MEOD Maximum Extended Operating Domain MSIVOOS Main Steam Isolation Valve Out-of-Service NCL Natural Circulation Line 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)

PRNM Power Range Neutron Monitoring (System)

PROOS Pressure Regulator Out-of-Service RBM Rod Block Monitor (Subsystem)

RDF Rated Drive Flow RFWT Reduced Feedwater Temperature RPT Recirculation Pump Trip RTP Rated Thermal Power SAD Amplitude Discriminator Setpoint (DSS-CD)

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

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 8, Revision 1 CAUTION References to COLR Figures or Tables should be made using titles only; Figure and Table numbers may change from cycle to cycle or between revisions.

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, TS 3.2.2 LCO (MCPR) 9, 10, 11, 12, TS 3.4.1 LCO (Recirculation loops 13, 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, TS 3.2.3 LCO (LHGR) 7, 8, 9, 10, TS 3.4.1 LCO (Recirculation loops 12, 13, 20 operating)

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

4. The Manual Backup Stability 18, 19 TS Table 3.3.1.1-1, Function 2.f Protection (BSP) Scram Region (OPRM Upscale)

(Region I), Manual BSP Controlled Entry Region (Region II), the modified TS 3.3.1.1, Condition I and J (Alternate Average Power Range Monitor instability detection)

(APRM) Simulated Thermal Power -

High Scram setpoints used in the Automated BSP Scram Region, the BSP Boundary for TS 3.3.1.1.

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

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 DSS-CD stability ranges.

The generation of this COLR is documented in Reference 30 and is based on analysis results documented in References 23, 27-29.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 9, Revision 1 APLHGR Limits Steady-state MAPLHGRSS limits are provided for Framatome Fuel (Table 16). These steady-state MAPLHGRSS limits must be modified as follows:

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

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.

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

MAPLHGR LimitTLO = MAPLHGRSS MAPLHGR LimitSLO = MAPLHGRSS x MAPFACSLO where MAPFACSLO = 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.

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

Power-dependent MCPRP 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.

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

The power-dependent MCPRp 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 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 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.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 10, Revision 1 LHGR Limits Steady-state LHGRSS limits are provided for Framatome Fuel (Table 10). These steady-state LHGRSS limits must be modified as follows:

Framatome Fuel LHGR limits have a core power and core flow dependency. Framatome Fuel power-dependent LHGRFACp multipliers (Tables 11-14) and flow-dependent LHGRFACf multipliers (Table 15) must be used to modify the steady-state LHGRSS limits (Table 10) for off-rated conditions.

Framatome Fuel power-dependent LHGRFACp 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.

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 = LHGRSS x (LHGRFACp, LHGRFACf)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.

The cycle-specific off-rated flow dependent LHGR set down bounds those assumed in the MELLLA+

plant-specific ECCS-LOCA analyses.

CDA Setpoints Brunswick Unit 2 has implemented the stability Detect and Suppress Solution - Confirmation Density (DSS-CD) solution using the Oscillation Power Range Monitor (OPRM) as described in Reference 19.

Plant-specific analyses for the DSS-CD Solution are provided in Reference 23. The Detect and Suppress function of the DSS-CD solution based on the OPRM system relies on the Confirmation Density Algorithm (CDA), which constitutes the licensing basis. The Backup Stability Protection (BSP) solution may be used by the plant in the event the OPRM Upscale function is declared inoperable.

The CDA enabled through the OPRM system and the BSP solution described in Reference 23 provide the stability licensing bases for B2C23. The safety evaluation report for Reference 19 concluded that the DSS-CD solution is acceptable subject to certain cycle-specific limitations and conditions. These cycle-specific limitations and conditions are met for B2C23.

A reload DSS-CD evaluation has been performed in accordance with the licensing methodology described in Reference 19 to confirm the Amplitude Discriminator Setpoint (SAD) of the CDA established in Reference 23. The Cycle 23 DSS-CD evaluation demonstrates that: 1) the DSS-CD Solution is applicable to B2C23; and, 2) the SAD value of 1.10 established in Reference 23 is confirmed for operation of B2C23.

The SAD setpoint value of 1.10 is applicable to TLO and to SLO.

Reference 19 describes two BSP options that are based on selected elements from three distinct constituents: BSP Manual Regions, BSP Boundary, and Automated BSP (ABSP) setpoints.

The Manual BSP region boundaries and the BSP Boundary were calculated for Brunswick Unit 2 Cycle 23 for nominal feedwater temperature operation and reduced feedwater temperature. The endpoints of the regions are defined in Table 3.1 and Table 3.2. The Manual BSP region boundary endpoints are

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 11, Revision 1 calculated with the Reference 18 methodology and connected using the Generic Shape Function (GSF),

which is described in Reference 19.

The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are determined for Cycle 23 and are defined in Table 3.3. These ABSP setpoints are applicable to both TLO and SLO as well as nominal and reduced feedwater temperature operation.

The Manual Backup Stability Protection (BSP) Regions I and II are documented on the Power/Flow maps as is the modified APRM Simulated Thermal Power (STP) high SCRAM setpoints and the BSP Boundary.

The power/flow maps (Figures 1-6) were validated for B2C23 based on Reference 29 using the Reference 19 methodology to facilitate operation under DSS-CD as implemented by Function 2.f of Table 3.3.1.1-1 and LCO Conditions I and J 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 23% RTP and below 75% drive flow (correlated to core flow) where the OPRM 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 there is increased probability the OPRM system may generate a scram to avoid an instability event. Figures 2, 4, and 6 support an inoperable OPRM and highlight the Manual Backup Stability Regions I and II, the modified APRM STP high SCRAM setpoints, and the BSP Boundary. 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-217345 which 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.

Base Case Operation

PROOS

SLO

Combined PROOS and TBVOOS

TBVOOS

Combined PROOS and FHOOS

FHOOS

Combined PROOS, TBVOOS and FHOOS

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.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 12, Revision 1

Any 1 inoperable SRV

2 inoperable TBV (Note that for TBVOOS, TBVOOS/FHOOS, PROOS/TBVOOS and PROOS/TBVOOS/FHOOS all 10 TBVs are assumed inoperable)

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, TBVOOS, PROOS and PROOS/TBVOOS.

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.

These power and flow limitations also apply when operating with jet pump loop flow mismatch conditions (LCO 3.4.1). The following must be considered when operating in SLO:

SLO is not permitted with RFWT (FHOOS).

SLO is not permitted with TBVOOS.

SLO is not permitted with MSIVOOS.

SLO is not permitted within the MELLLA+ operating domain.

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).

When in SLO, Figures 3 and 4 implement the corrective action for AR-217345 which 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.

If OPRMs are inoperable in SLO, the expansion of the ABSP region results in power being restricted to 39% RTP as shown in Figure 4.

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 and in the MELLLA+ domain after achieving 13 GWd/MTU cycle exposure 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.

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

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

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

TBVOOS operation coincident with FHOOS and PROOS is supported using the combined

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 13, Revision 1 PROOS/TBVOOS/FHOOS limits.

SLO is not permitted with TBVOOS.

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:

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 >10F and reactor power 23% RTP requires use of the 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 DSS-CD Power/Flow Maps (Figures 5 and 6) must be used.

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

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

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

SLO is not permitted with RFWT.

FWTR operation within the MELLLA+ operating domain is not allowed.

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 and in the MELLLA+ domain after achieving 13 GWd/MTU cycle exposure with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. The following must be considered when operating with PROOS:

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.

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.

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.

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

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

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

MELLLA+ Implementation Revision 1 of this COLR is being issued to support implementation of B2C23 operation in the MELLLA+

domain.

Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-1325 B2C23 Core Operating Limits Report, BNEI-0400-0005-001 Page 14, Revision 1 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.