RA-19-0162, Cycle 24 Core Operating Limits Report (COLR)
| ML19079A038 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/19/2019 |
| From: | Pierce J Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-19-0162 | |
| Download: ML19079A038 (40) | |
Text
Brunswick Nuclear Plant
( -, DUKE P.O. Box 10429 ENERGY Southport, NC 28461 March 19, 2019 Serial: RA-19-0162 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 No. 50-324 Unit 2 Cycle 24 Core Operating Limits Report (COLA)
Reference Letter from Mark McPherson (Duke Energy) to NRC Document Control Desk, Unit 2 Cycle 23 Core Operating Limits Report (COLR), dated April 9, 2017, ADAMS Accession Number ML17100A840 Ladies and Gentlemen:
Enclosed is a copy of the Core Operating Limits Report (COLA) for Brunswick Steam Electric Plant (BSEP), Unit 2 Cycle 24 operation. Duke Energy Progress, LLC (Duke Energy), is providing the enclosed COLA in accordance with Brunswick Unit 2 Technical Specification 5.6.5.d. The enclosed COLA supersedes the report previously submitted by letter dated April 9, 2017 (i.e., Reference).
This letter and the enclosed COLA do not contain any regulatory commitments.
Please refer any questions regarding this submittal to Mr. Mark Turkal, Lead Licensing Engineer, at {910) 832-3066.
Sincerely, ~
Jer Pierce Manager - Nuclear Support Services Brunswick Steam Electric Plant MAT/mat
Enclosure:
Brunswick Unit 2, Cycle 24 Core Operating Limits Report
U.S. Nuclear Regulatory Commission Page 2 of 2 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. Gale Smith, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U.S. Nuclear Regulatory Commission ATTN: Mr. Dennis J. Galvin (Mail Stop OWFN 8B1A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission (Electronic Copy Only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net
RA-19-0162 Enclosure Brunswick Unit 2, Cycle 24 Core Operating Limits Report
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Cale. No. 2821-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEl-0400-0028 Rev. 0 Page 1 BRUNSWICK UNIT 2, CYCLE 24 CORE OPERATING LIMITS REPORT March 2019 Prepared by:
Steve Evans Brunswick Nuclear Design Reviewed by:
R nWells Brunswick Nuclear Design i88935 (326946) dc=com, dc=duke-energy, d c=ent, dc=nam, ou= Accounts, ou=Personal, ou=PNTransitional, cnai88935 [326946), emailKAllen.Butler@duke-energy.com Site Inspection by: 2019.03.1913:11:21-04'00' Allen Butler Brunswick Reactor Engineering Approved by:
~J449 Manager, Brunswick Nuclear Design
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 2 LIST OF EFFECTIVE PAGES Page(s) Revision 1- 37 0 This document consists of 37 total pages.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 References..................................................................................................................................... 14
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 2 Cycle 24 Transient Analysis ................................ 19 Table 5: Power-Dependent MCPRp Limits .................................................................................. 20 NSS Insertion Times - BOC to < EOCLB Table 6: Power-Dependent MCPRp Limits .................................................................................. 21 TSSS Insertion Times - BOC to < EOCLB Table 7: Power-Dependent MCPRp Limits .................................................................................. 22 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)
Table 8: Power-Dependent MCPRp Limits .................................................................................. 23 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)
Table 9: Flow-Dependent MCPRf Limits ..................................................................................... 24 Table 10: Framatome Fuel Steady-State LHGRSS Limits .............................................................. 25 Table 11: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 26 NSS Insertion Times - BOC to < EOCLB Table 12: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 27 TSSS Insertion Times - BOC to < EOCLB Table 13: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 28 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)
Table 14: Framatome Fuel Power-Dependent LHGRFACp Multipliers .......................................... 29 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)
Table 15: Framatome Fuel Flow-Dependent LHGRFACf Multipliers ............................................. 30 Table 16: Framatome Fuel Steady-State MAPLHGRSS Limits ...................................................... 31
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 ............................................................................... 32 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability DSS-CD Power/Flow Map ............................................................................... 33 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability DSS-CD Power/Flow Map ............................................................................... 34 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability DSS-CD Power/Flow Map ............................................................................... 35 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability DSS-CD Power/Flow Map ............................................................................... 36 OPRM Operable, FWTR, 2923 MWt Figure 6: Stability DSS-CD Power/Flow Map ............................................................................... 37 OPRM Inoperable, FWTR, 2923 MWt
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 LTA Lead Test Assembly LTSP Low Trip Set Point
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 2, Cycle 24 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 )
- 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 B2C24, 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 9 APLHGR Limits Steady-state MAPLHGRSS limits are provided for Framatome Fuel (Table 16). These steady-state MAPLHGRSS limits must be modified as follows:
- 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. 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 10 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 B2C24. 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 B2C24.
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 24 DSS-CD evaluation demonstrates that: 1) the DSS-CD Solution is applicable to B2C24; and, 2) the SAD value of 1.10 established in Reference 23 is confirmed for operation of B2C24.
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 validated for Brunswick Unit 2 Cycle 24 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 11 The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are determined for Cycle 24 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 B2C24 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-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 24 is analyzed for the following operating conditions with applicable MCPR, APLHGR and LHGR limits.
- Base Case Operation
- 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 12
- Any 1 inoperable SRV
- 2 inoperable TBV (Note that for TBVOOS, TBVOOS/FHOOS, all 10 TBVs are assumed inoperable)
- Up to 50% of the LPRMs OOS Please note that during FFTR/Coastdown, FHOOS is included in Base Case Operation and TBVOOS.
Single Loop Operation Brunswick Unit 2, Cycle 24 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 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-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.
- 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 24 may operate with an inoperable Main Turbine Bypass System over the entire MEOD range and in the MELLLA+ domain for all cycle exposures with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. An operable Main Turbine Bypass System with only two 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:
- 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 combined TBVOOS/FHOOS limits.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 13 Feedwater Temperature Reduction Brunswick Unit 2, Cycle 24 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.
- 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.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 14 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, 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 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 4, September 2018.
- 28. BNP Design Calculation 0B21-2045, BNP Power/Flow Maps For MELLLA+, Revision 1, September 2017.
- 29. ANP-3742P, Brunswick Unit 2 Cycle 24 Reload Safety Analysis, Revision 0, December 2018.
- 30. BNP Design Calculation 2B21-2067, Preparation of the B2C24 Core Operating Limits Report, Revision 0, March 2019.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 17 Table 2 RBM Operability Requirements 2 IF the following conditions are met, THEN RBM Not Required Operable Thermal Power
(% rated) MCPR 1.83 TLO 29% and < 90%
1.86 SLO 90% 1.51 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 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 %RTP/%RDF 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 19 Table 4 Exposure Basis 6 for Brunswick Unit 2 Cycle 24 Transient Analysis Core Average Exposure Comments (MWd/MTU)
Breakpoint for design basis rod patterns to 36,575 EOFP + 15 EFPD (NEOC/EOCLB 7)
End of cycle with FFTR/Coastdown -
38,157 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.
7 NEOC exposure for Unit 2 Cycle 24 is defined as the same as the EOCLB exposure.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 20 Table 5 Power-Dependent MCPRp Limits 8 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.34 1.45 80.0 1.44 1.47 50.0 1.66 1.60 Base
> 65%F 65%F > 65%F 65%F Case 50.0 1.83 1.74 2.02 1.92 Operation 26.0 2.22 2.11 2.42 2.33 26.0 2.24 2.14 2.44 2.36 23.0 2.32 2.22 2.50 2.44 100.0 1.36 1.47 80.0 1.44 1.51 50.0 1.67 1.61
> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.83 1.74 2.02 1.92 26.0 2.22 2.11 2.42 2.33 26.0 2.72 2.60 3.04 2.94 23.0 2.87 2.77 3.20 3.16 100.0 1.34 1.45 80.0 1.44 1.47 50.0 1.67 1.60
> 65%F 65%F > 65%F 65%F FHOOS 50.0 1.83 1.74 2.02 1.92 26.0 2.22 2.11 2.42 2.33 26.0 2.24 2.14 2.44 2.36 23.0 2.32 2.22 2.50 2.44 100.0 1.36 1.47 80.0 1.44 1.51 TBVOOS 50.0 1.67 1.64 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.83 1.74 2.02 1.92 26.0 2.22 2.11 2.42 2.33 26.0 2.80 2.71 3.18 3.10 23.0 2.97 2.88 3.33 3.30 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.
FHOOS not permitted in the MELLLA+ domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 21 Table 6 Power-Dependent MCPRp Limits 9 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp 100.0 1.37 1.48 80.0 1.44 1.49 50.0 1.66 1.60 Base
> 65%F 65%F > 65%F 65%F Case 50.0 1.85 1.76 2.03 1.93 Operation 26.0 2.23 2.13 2.43 2.35 26.0 2.24 2.14 2.44 2.36 23.0 2.32 2.22 2.50 2.44 100.0 1.39 1.50 80.0 1.44 1.54 50.0 1.67 1.63
> 65%F 65%F > 65%F 65%F TBVOOS 50.0 1.85 1.76 2.03 1.93 26.0 2.23 2.13 2.43 2.35 26.0 2.72 2.60 3.04 2.94 23.0 2.87 2.77 3.20 3.16 100.0 1.37 1.48 80.0 1.44 1.49 50.0 1.67 1.60
> 65%F 65%F > 65%F 65%F FHOOS 50.0 1.85 1.76 2.03 1.93 26.0 2.23 2.13 2.43 2.35 26.0 2.24 2.14 2.44 2.36 23.0 2.32 2.22 2.50 2.44 100.0 1.39 1.50 80.0 1.44 1.54 TBVOOS 50.0 1.67 1.66 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 1.85 1.76 2.03 1.93 26.0 2.23 2.13 2.43 2.35 26.0 2.80 2.71 3.18 3.10 23.0 2.97 2.88 3.33 3.30 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.
FHOOS not permitted in the MELLLA+ domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 22 Table 7 Power-Dependent MCPRp Limits 10 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp Base Case 100.0 1.36 1.45 Operation 80.0 1.44 1.47 50.0 1.67 1.60 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 1.83 1.74 2.02 1.92 26.0 2.22 2.11 2.42 2.33 (Bounds operation 26.0 2.24 2.14 2.44 2.36 with NFWT) 23.0 2.32 2.22 2.50 2.44 TBVOOS 100.0 1.37 1.47 80.0 1.44 1.51 (FFTR/FHOOS 50.0 1.67 1.64 included) > 65%F 65%F > 65%F 65%F 50.0 1.83 1.74 2.02 1.92 (Bounds operation 26.0 2.22 2.11 2.42 2.33 with NFWT) 26.0 2.80 2.71 3.18 3.10 23.0 2.97 2.88 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.
FHOOS not permitted in the MELLLA+ domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 23 Table 8 Power-Dependent MCPRp Limits 11 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 10XM ATRIUM 11 LTA Condition (% rated) MCPRp MCPRp Base Case 100.0 1.39 1.48 Operation 80.0 1.44 1.49 50.0 1.67 1.60 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 1.85 1.76 2.03 1.93 26.0 2.23 2.13 2.43 2.35 (Bounds operation 26.0 2.24 2.14 2.44 2.36 with NFWT) 23.0 2.32 2.22 2.50 2.44 TBVOOS 100.0 1.39 1.50 80.0 1.44 1.54 (FFTR/FHOOS 50.0 1.67 1.66 included) > 65%F 65%F > 65%F 65%F 50.0 1.85 1.76 2.03 1.93 (Bounds operation 26.0 2.23 2.13 2.43 2.35 with NFWT) 26.0 2.80 2.71 3.18 3.10 23.0 2.97 2.88 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.
FHOOS not permitted in the MELLLA+ domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 24 Table 9 Flow-Dependent MCPRf Limits 12 Core Flow ATRIUM 10XM ATRIUM 11 LTA
(% of rated) MCPRf MCPRf 13 0.0 1.57 1.80 31.0 1.57 1.80 55.0 1.49 --
80.0 1.30 --
100.0 1.30 1.20 107.0 1.30 1.20 12 Limits valid for all SCRAM insertion times, all core average exposure ranges, all EOOS scenarios, and both TLO
& SLO.
13 -- indicates that this fuel type does not have a breakpoint at the indicated exposure
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 25 Table 10 Framatome Fuel Steady-State LHGRSS Limits Peak ATRIUM 10XM ATRIUM 11 LTA Pellet Exposure LHGR LHGR 14 (GWd/MTU) (kW/ft) (kW/ft) 0.0 14.1 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
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 26 Table 11 Framatome Fuel Power-Dependent LHGRFACp Multipliers 15 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 1.00 0.92 Base > 65%F 65%F > 65%F 65%F Case Operation 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.65 0.79 0.64 0.66 23.0 0.62 0.73 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 1.00 0.92
> 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.44 0.51 0.43 0.50 23.0 0.41 0.45 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 1.00 0.92
> 65%F 65%F > 65%F 65%F FHOOS 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.65 0.76 0.64 0.66 23.0 0.62 0.73 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 TBVOOS 50.0 1.00 0.92 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.41 0.46 0.40 0.46 23.0 0.39 0.42 0.38 0.43 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. FHOOS not permitted in the MELLLA+
domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 27 Table 12 Framatome Fuel Power-Dependent LHGRFACp Multipliers 16 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 1.00 0.92 Base > 65%F 65%F > 65%F 65%F Case Operation 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.65 0.79 0.64 0.66 23.0 0.62 0.73 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 1.00 0.92
> 65%F 65%F > 65%F 65%F TBVOOS 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.44 0.51 0.43 0.50 23.0 0.41 0.45 0.40 0.46 100.0 1.00 1.00 90.0 1.00 1.00 50.0 1.00 0.92
> 65%F 65%F > 65%F 65%F FHOOS 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.65 0.76 0.64 0.66 23.0 0.62 0.73 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 TBVOOS 50.0 1.00 0.92 and > 65%F 65%F > 65%F 65%F FHOOS 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 26.0 0.41 0.46 0.40 0.46 23.0 0.39 0.42 0.38 0.43 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. FHOOS not permitted in the MELLLA+
domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 28 Table 13 Framatome Fuel Power-Dependent LHGRFACp Multipliers 17 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 1.00 0.92 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 (Bounds operation 26.0 0.65 0.76 0.64 0.66 with NFWT) 23.0 0.62 0.73 0.60 0.64 TBVOOS 100.0 1.00 1.00 90.0 1.00 1.00 (FFTR/FHOOS 50.0 1.00 0.92 included) > 65%F 65%F > 65%F 65%F 50.0 0.88 0.98 0.86 0.86 (Bounds operation 26.0 0.65 0.79 0.64 0.66 with NFWT) 26.0 0.41 0.46 0.40 0.46 23.0 0.39 0.42 0.38 0.43 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. FHOOS not permitted in the MELLLA+
domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 29 Table 14 Framatome Fuel Power-Dependent LHGRFACp Multipliers 18 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 1.00 0.92 (FFTR/FHOOS > 65%F 65%F > 65%F 65%F included) 50.0 0.88 0.98 0.86 0.86 26.0 0.65 0.79 0.64 0.66 (Bounds operation 26.0 0.65 0.76 0.64 0.66 with NFWT) 23.0 0.62 0.73 0.60 0.64 TBVOOS 100.0 1.00 1.00 90.0 1.00 1.00 (FFTR/FHOOS 50.0 1.00 0.92 included) > 65%F 65%F > 65%F 65%F 50.0 0.88 0.98 0.86 0.86 (Bounds operation 26.0 0.65 0.79 0.64 0.66 with NFWT) 26.0 0.41 0.46 0.40 0.46 23.0 0.39 0.42 0.38 0.43 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. FHOOS not permitted in the MELLLA+
domain.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 30 Table 15 Framatome Fuel Flow-Dependent LHGRFACf Multipliers 19 ATRIUM 10XM and Core Flow ATRIUM 11 LTA
(% of rated) LHGRFACf 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 19 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 31 Table 16 Framatome Fuel Steady-State MAPLHGRSS Limits20, 21 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 20 Framatome Fuel MAPLHGR limits do not have a power, flow, or EOOS dependency.
21 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.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 32 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 lit ii 111 I 111 I 11 I,_ I I I I I I I 85 84 59.50 58.43 49.97 48.94 80.69 80.79
% Power W-- j_j_j_,-i-- I -1---1-- __j_j_, ~ * -i-+-+-++-H-1+-l 11 +++++l-+-++-11>-+-H+*~* _j__j_j_,
11-++ ++-IH- .j_j_j_. 83 57.37 47.90 80.90
+-
60.0 82 56.31 46.87 81.05 81 55.25 45.84 81.21 111 ttt° 80 54.20 44.81 81.36
,...,__ ~
+++---H-'-++-4 R
.j_j_j_. 79 53.16 43.77 81.51 78 52.12 42.74 81.67 50.0 MELLLA Line and I e 77 51.08 --- 81.82 BSP Boundary SLO Entry Rod Line tf-t---t-H-+-l+/-+/-t C gI fil 76 75 50.05 49.02 81.98 82.13 111 ttt° 74 48.00 --- 82.29 40.0 F i 30.0
- ._1 Scram
_ - - Avoidance IRegion T - _'"'T'" - - - - 'lf ti --tl
.b-
_//i__.. - **
I l
oI Ill TTr nI ~ I l l I l l 73 72 71 70 69 46.98 45.96 44.95 43.94 42.94 82.44 82.60 82.75 82.91 83.06 68 41.94 --- 83.22 67 40.95 --- 83.37 66 39.96 --- 83.52 20.0 65 38.97 --- 83.68
=m-=11 111 I* 111 111 64 63 37.99 37.01 83.83 83.99 10.0
+-+-+-
Natural Circulation --
~
--1,--1++++++t++l++++++1..... 1
~
I OPRM Enabled Region
_j__j_j_, .j_j_j_.
62 61 60 36.04 35.06 34.10 84.14 84.30 84.45
-+/-tt- Line 35% Approximate 11 111 111 59 33.13 --- 84.61 0.0 1Tr 11 Minimum Pump Speed Minimum Power Line I 58 32.17 --- 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-2045, Revision 1
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 33 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
'ttt" tt m 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
+-rTT1 TT I I ABSP STP Rod Block T -+- r1 t t=--r . J/ '
89 88 63.83 62.74 54.10 53.06 80.47 80.47 70.0 t--1---W-a 1-- +-IIf-* * -......- -......- - ' -I- W- I ~-1' 87 86 61.66 60.58 52.03 51.00 80.51 80.60 I I I I I I I I , I I I I I I 85 59.50 49.97 80.69 84 58.43 48.94 80.79
% Power
,-.J..j..j., 1-- W-- * -i-+-+-++-H-1+-l II +-H-++l-+-++-11>-+-H+*~* _j__j_j_,
11~14+-IH- .1-)..j_. 83 57.37 47.90 80.90
+-
82 56.31 46.87 81.05 60.0 81 55.25 45.84 81.21 111 ttt° 80 54.20 44.81 81.36
+-H----H-'-++-4 .1-)..j_. 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
_, I . 1---.LL
/ I .,,' r-n-1-rrr 68 41.94 --- 83.22 Operator Awareness / - Ill Ill 67 40.95 --- 83.37 66 39.96 --- 83.52 I I 65 38.97 --- 83.68 20.0 64 37.99 --- 83.83 111 111 I* 111 111 63 37.01 --- 83.99 10.0 f-i_.j__j_,
+-+-+-
Natural Circulation -- .
~ --1,--1++++++t++l++++++1.....
~
1 I OPRM Enabled Region
_j__j_j_, .1-)..j_.
62 61 60 36.04 35.06 34.10 84.14 84.30 84.45
-+/-tt- Line 35% Approximate 11 111 111 59 33.13 --- 84.61 0.0 1Tr 11 Minimum Pump Speed Minimum Power Line I 58 32.17 --- 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-2045, Revision 1
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 34 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 I I I I I I I I U..L -~ ~ ! , I I I I I I 85 84 59.50 58.43 49.97 48.94 80.69 80.79
% Power
,-.J..j..j.,
+-
1-- W-- -1-+- -1-1..,.,.
- 1 1-+-Hll-+-I+++++l-+++l+-1+-++r*~* _j__j_j_,
- -1+++++-11-+I++ .j_j_j_. 83 57.37 47.90 80.90 82 56.31 46.87 81.05 60.0 81 55.25 45.84 81.21
'tt"t" tt m-- t ttt" i ii L IT" ~ I 111 ttt"" 80 54.20 44.81 81.36
_~ _ _-1-1--
_...__ _ _-1-_ _ _ _j_j_j_,
W-- _ -i-- l-+-~ \ ~ '"
, *-+-+--++-
111-+l+-l-+-+-IH 11-+-+--++II++- +++---H-c.++-< .j_j_j_. 79 53.16 43.77 81.51 R ---+ 78 52.12 42.74 81.67 50.0 MELLLA Line and
- +I// /, ; I e 77 51.08 --- 81.82 1~7~71= --=1~ BSP Boundary
~ll~l~ ~rl~ -l~l ~l ~ 1 !,+t-++-,H-H ++t-+++++t-1++-+++---+-l+/-+/-t C g
~*-+++-H-++l-+-H-++--++-t-1-+-+---~---+-0 +-
llf-HI fil m--
76 75 50.05 49.02 81.98 82.13 74 48.00 --- 82.29 40.0
- F i 73 46.98 --- 82.44 I ** o 72 45.96 --- 82.60 TTr 71 44.95 --- 82.75 Scram Avoidance Region I ' l Ill n ~ 111 111 70 43.94 --- 82.91 30.0 r 69 42.94 --- 83.06 68 41.94 --- 83.22 111 111 67 40.95 --- 83.37
_j__j_j_, .j_j_j_. 66 39.96 --- 83.52 20.0 65 38.97 --- 83.68 111 111 I*
". 45 Mlb/hr Max Core Flow
+----+-1
- ~-
'---+++-----+++---1---+l-+-
I +---
l +-+
ll +-
1-
- 64 37.99 --- 83.83 63 37.01 --- 83.99 Natural 62 36.04 --- 84.14 10.0 rH Circulation I OPRM Enabled Region 61 35.06 --- 84.30 60 34.10 --- 84.45
-+/-tt- Line 35% Approximate 11 111 111 59 33.13 --- 84.61 0.0 1Tr 11 Minimum Pump Speed Minimum Power Line I 58 32.17 --- 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-2045, Revision 1
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 35 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 3.3 120.0 110.0 ir=t +t-t+-----t ilL 1 t+t+-+
UL .L Fi= MELLLA+ Line (MELLLA+ Operations prohibited during SLO) lf =U1---+/-1 APRM STP Scram I +t++ ...i,.i+-
~ UL Power Minimum Minimum Maximum (MELLLA) (MELLLA+)
Core Flow Core Flow (ICF)
Core Flow t--1---W-a 1-- ilL 1---W-a 1+ --1-1-- __j_j_, ,-.1...).,
-il+~ >-+-l-1+ --1-1-- ilL --1-- 1- a y . . >--+--i----1--,
J,),Y l-1-)_. --
100 99 Mlbs/hr 76.19 75.04 Mlbs/hr 65.45 64.42 Mlbs/hr 80.47 80.47 100.0 90.0
,-.J...Lj.,
+--ttt7
'ttt"
.L.L tt tt I
ilL iTT*
111 ttt' 1
t t
I UL .LI-.L.
tit7 i i, ttt" *ti-1
.L.L.L tt,*
--.L).., +1--.l, it' *+-tt7 u
II ABSP STP Scram 11 UL iTT 111
.L.L" ti 11 I
I~ ~
A -~
~/n1 -
~ -~
- -+----
r
~
~
+. -
APRM STP Rod Block TIT' TTr 98 97 96 95 94 93 92 73.89 72.75 71.61 70.49 69.36 68.25 67.13 63.39 62.35 61.32 60.29 59.26 58.22 57.19 80.47 80.47 80.47 80.47 80.47 80.47 80.47 80.0 ..... \- ** 91 90 66.03 64.93 56.16 55.13 80.47 80.47
+-rTT1 TT ,TT T tit7 lffi-- ~ t..... TIT1 TTr 89 63.83 54.10 80.47
'V II II ABSP STP Rod Block 88 62.74 53.06 80.47 t--1---W-a 1-- ilL 1---W-a -1-+< liL 87 61.66 52.03 80.51 I ~
86 60.58 51.00 80.60 70.0 111 I 111 I 111 W-.' 7 I*-
I I ' 111 111 85 84 59.50 58.43 49.97 48.94 80.69 80.79 II
% Power
,-.J...Lj., .L.L ilL 1 UL .LI-.L. .L.L.L .L.L.L 60.0
+- I I. 111 111 11
'~ _j_j_j_,
83 82 57.37 56.31 47.90 46.87 80.90 81.05
,,..... \
81 55.25 45.84 81.21 tt -m--
j 'I
'ttt" t ttt" i" I iTT 50.0
,-.J...Lj., .L.L ilL MELLLA Line and 1 UL .L
~7) \
I I
I 111 111 R *--+
I I I
.L.L.L 80 79 78 77 54.20 53.16 52.12 51.08 44.81 43.77 42.74 81.36 81.51 81.67 81.82 1 It
./',- ./i.
BSP Boundary I e
+/-it iii 76 50.05 --- 81.98 40.0 l Region I - Manual Scram I
~/"T *I C g F i 111 iTT 75 74 73 49.02 48.00 46.98 82.13 82.29 82.44 I
I ~ I ** .Ht--+
o 72 45.96 --- 82.60 1Region II - Controlled Entry I / JI I I
I n ~
111 TTr 71 70 44.95 43.94 82.75 82.91 l 111 111 30.0 r 69 42.94 --- 83.06
_, I . 1---1...L
/ 1 ~ TIT1 TTr 68 41.94 --- 83.22
! -~
Operator Awareness 111 111 67 40.95 --- 83.37
,trrrt,.,...,... 111'.-rt: .... --.L).., _j_j_j_, .L.L.L 66 39.96 --- 83.52 I I 1 --,-.
65 38.97 --- 83.68 20.0 111 111 I*
". 45 Mlb/hr Max Core Flow
+-------+1- -
-r 111 111 UL 64 63 37.99 37.01 83.83 83.99
...UL
--fil"* .L.L.
I _j_j_j_,
Natural ....UL_ 62 36.04 --- 84.14 10.0
+-+-+-
Circulation
- -....._ I OPRM Enabled Region
, 1 1-ll-61 35.06 --- 84.30 60 34.10 --- 84.45
-+/-tt- Line Jib ] 35% Approximate it L 11 111 111 59 33.13 --- 84.61 w
I A
~ Minimum Pump Speed ~ 11 Minimum Power Line I 58 32.17 --- 84.70 I
11 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 36 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 3.3 120.0 110.0 ir=t +t-t+-----t ilL 1 t+t+-+
UL .L Fi= MELLLA+ Line (MELLLA+ Operations prohibited during FWTR) lf ++ f---r i l i.,,,,,,,,,,- '-4 L L .L APRM STP Scram I 111 111 Power Minimum Minimum Maximum (MELLLA) (MELLLA+)
Core Flow Core Flow (ICF)
Core Flow fil tt ilL 1---W-a 1+ --1---1-- __j_j_, ,-.1...).,
-il+~ l?,*- ~
--1-- I- +a -JI APRM STP Rod Block I - %
99 Mlbs/hr 100 -76.19 75.04 Mlbs/hr 65.45 64.42 Mlbs/hr 80.47 80.47 100.0 98 73.89 63.39 80.47 Jit ii L ~ ), ~ _. ab- -l-L 97 72.75 62.35 80.47
~ ~ ~ ,Jtt ilL 1 UL .LI-+-- .LW-. e-1-L .LW-.
tt tt TTT*
111 t
I tit7 i i,
~
tit'*
),L
~/n
- 1 I II"'
"7 ftf-:- tit7 TTT 111 96 95 71.61 70.49 61.32 60.29 80.47 80.47 90.0 Ill -
94 69.36 59.26 80.47 t m- 4"tt< ~ ii ett.A 80.0 1ff IT ttt' ttt" 1 1
'I' - I ttt -
~ ; l ttt' Iii 93 92 91 90 68.25 67.13 66.03 64.93 58.22 57.19 56.16 55.13 80.47 80.47 80.47 80.47
~
TIT ilL T
tit7 i 1---W-a r/ nr,,,
y
~ TT1
__j_j_, +1--1--a TTT'i: ~
1-)... 1 - ~
' C---- tit7 TTT
---1---1--a 1-)...
89 88 87 63.83 62.74 61.66 54.10 53.06 52.03 80.47 80.47 80.51 70.0 ,, 86 60.58 51.00 80.60 fil tt 111 I 111 I 7
11 1- ,
~ -l-L 111 111 85 84 59.50 58.43 49.97 48.94 80.69 80.79
% Power ilL 1 LLL .L I J,.L.L .LW-.
,. L j 111 111 11 83 82 57.37 56.31 47.90 46.87 80.90 81.05 60.0 J[ IT TTT ilL t
1 ttt" i' LLL .L ltt'
__y ......... ' I \ ...J.....
- 111 Iii
.LW-.
81 80 79 55.25 54.20 53.16 45.84 44.81 43.77 81.21 81.36 81.51 y'
~
R *--+ 78 52.12 42.74 81.67 50.0 MELLLA Line and 77 51.08 --- 81.82
,~ +/-it iii I e
~
1 BSP Boundary SLO Entry 76 50.05 --- 81.98 L-----'"" ------
I fl Rod Line C gI Iii 75 49.02 --- 82.13
=Hf tt 111 111
~
111
~ 74 48.00 --- 82.29 40.0 (SLO prohibited F i 73 46.98 --- 82.44 I ,H4 30.0 1 Scram Avoidance Region If ~
?{-- ---- during FWTR) - oI nI ~
111 111 TTT 111 72 71 70 69 45.96 44.95 43.94 42.94 82.60 82.75 82.91 83.06 r
TTT t tit7 i I ,: tit7 TTT 68 41.94 --- 83.22
~ - 111 I 111 - I 111 111 67 40.95 --- 83.37 ilL 1 LLL .L ,trrrt,.,...,... .... L -l-L .LW-. 66 39.96 --- 83.52 20.0 '"'" +----+J_.....
65 38.97 --- 83.68
' 64 37.99 --- 83.83
=m-=11 111
--fil-"*
I*
I * .Ll...
-rI 111
-l-L .LW-.
111 63 37.01 --- 83.99 Natural .....LW-.- 62 36.04 --- 84.14 Hr :;-....._ I OPRM Enabled Region 1-ll-61 35.06 --- 84.30 10.0 Circulation
-J+/-t--- Line Jc:;- ] Minimum 35% Approximate it L
1 11 111 111 60 59 34.10 33.13 84.45 84.61 w
I A Pump Speed ~ 11 Minimum Power Line 58 32.17 --- 84.70 I - 'TT' I 11 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. 2B21-2067 Rev. 0 B2C24 Core Operating Limits Report, BNEI-0400-0028 Rev. 0 Page 37 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 3.3 120.0 110.0 ir=t +t-t+-----t ilL 1 t+t+-+
UL .L Fi= MELLLA+ Line (MELLLA+ Operations prohibited during FWTR) lf ++ f---r i l i.,,,,,,,,,,- '-4 L L .L APRM STP Scram I 111 111 Power Minimum Minimum Maximum (MELLLA) (MELLLA+)
Core Flow Core Flow (ICF)
Core Flow t--1---W-a 1-- ilL 1---W-a - I+ --1-1-- __j_j_, ,-.1...).,
-il+~ l?,*- ~
--1-- I- +a -JI APRM STP Rod Block I 100 99 Mlbs/hr 76.19 75.04 Mlbs/hr 65.45 64.42 Mlbs/hr 80.47 80.47 100.0 98 73.89 63.39 80.47 u L ~ .... e-1-L ), ~ -- ab- -l-L 97 72.75 62.35 80.47
~ ~ ~ ,Jtt
+--1--.LL ilL 1 UL .L I-+-- U.L .LW-.
+--ttt7 tt I
it,*
111 t
I tit7 i I, tit'*
~
),L
~A
- 1 I "7
ftf-:- tit7 tt, 111 96 95 71.61 70.49 61.32 60.29 80.47 80.47 90.0 Ill -
.t 94 69.36 59.26 80.47
'ttt" tt m ~u+/- ~~~~~~
~
71 ........ tt" ~ 1 ett.A ~ttt -; I ttt" -m-- 93 92 91 68.25 67.13 66.03 58.22 57.19 56.16 80.47 80.47 80.47 80.0 70.0
+-rTT1 t--1---W-a TT 1--
11 11 1 I
ABSP STP Rod Block I-
' -I-T r/
n tt, ~ T
,~J - ;
W/1 -...1,..
' u:::'...._ tit7 t t ,
---1--l-a 1-)...
90 89 88 87 86 64.93 63.83 62.74 61.66 60.58 55.13 54.10 53.06 52.03 51.00 80.47 80.47 80.47 80.51 80.60
~,1/;I I I 85 59.50 49.97 80.69 111 I 111 I 111
~ -l-L 111 111 84 58.43 48.94 80.79
% Power u
60.0
+--1--.LL
+-
ilL 1 LLL .LI-+-- U.L
-Ullo/ 111 111 11
.LW-. 83 82 57.37 56.31 47.90 46.87 80.90 81.05
'ttt" f--i...LL tt u
ttt' ilL t
1 ttt" i' LLL .L fi'I I I \ ...J..... ~
- H--+
R I I 111 ttt°
.LW-.
81 80 79 78 55.25 54.20 53.16 52.12 45.84 44.81 43.77 42.74 81.21 81.36 81.51 81.67 50.0 MELLLA Line and 77 51.08 --- 81.82 i,~ +/-it iii I e 1 BSP Boundary I -1
/) SLO Entry 76 50.05 --- 81.98 40.0 l Region I - Manual Scram "T
- --I I
L L-----'"" ------ Rod Line (SLO prohibited C g F i I I
,H4 111 -m-- 75 74 73 49.02 48.00 46.98 82.13 82.29 82.44
~ / J...-
1Region II - Controlled Entry * ~ ,,,,,,,- I V during FWTR)
- o I I 111 tt, 72 71 45.96 44.95 82.60 82.75
- ~! -
n 70 43.94 --- 82.91 30.0 _, I
- ~
~
' I I ~ 111 tit7 t t ,
111 69 68 42.94 41.94 83.06 83.22 20.0 Operator Awareness I
-- ,trrrt,.,...,... 111',.---r-tt, 4= 111
-l-L .LW-.
111 67 66 65 40.95 39.96 38.97 83.37 83.52 83.68
" +-----+ - 64 37.99 --- 83.83 111 f--i...LL 111
---fil-"*
I*
I * .Ll...
-r I I 111
-l-L .LW-.
111 63 37.01 --- 83.99
~ _-I- 62 36.04 --- 84.14
+-+-+-
Natural :;-....._ I OPRM Enabled Region I. U.
61 35.06 --- 84.30 10.0 Circulation
-j+/-t---- Line Jc:;- ] Minimum 35% Approximate it L
1 11 111 111 60 59 34.10 33.13 84.45 84.61 w
I A .- Tr Pump Speed ~ 11 Minimum Power Line 58 32.17 --- 84.70 I I 11 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