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{{Adams
#REDIRECT [[BSEP 15-0022, Cycle 22 Core Operating Limits Report (COLR)]]
| number = ML15091A406
| issue date = 03/23/2015
| title = Cycle 22 Core Operating Limits Report (Colr)
| author name = Pope A H
| author affiliation = Duke Energy Carolinas, LLC, Duke Energy Corp
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000324
| license number = DPR-062
| contact person =
| case reference number = BSEP 15-0022
| document report number = 2B21-2020, Rev. 0
| document type = Fuel Cycle Reload Report, Letter
| page count = 42
}}
 
=Text=
{{#Wiki_filter:~ENERGY, Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 March 23, 2015 Serial: BSEP 15-0022 10 CFR 50.4 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 22 Core Operating Limits Report (COLR)
 
==Reference:==
 
Letter from Annette H. Pope (Duke Energy) to NRC Document Control Desk, Unit 2 Cycle 21 Core Operating Limits Report (COLR), dated April 23, 2013, ADAMS Accession Number ML1 3142A031 Ladies and Gentlemen:
Enclosed is a copy of the Core Operating Limits Report (COLR) for Brunswick Steam Electric Plant (BSEP), Unit 2 Cycle 22 operation.
Duke Energy Progress, Inc., is providing the enclosed COLR in accordance with Brunswick Unit 2 Technical Specification 5.6.5.d. The enclosed COLR supersedes the report previously submitted by letter dated April 23, 2013.This document contains no regulatory commitments.
Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager -Regulatory Affairs, at (910) 457-2487.Sincerely, Annette H. Pope Director -Organizational EffectiVeness Brunswick Steam Electric Plant WRM/wrm
 
==Enclosure:==
 
Brunswick Unit 2, Cycle 22 Core Operating Limits Report, March 2015 A 00 U.S. Nuclear Regulatory Commission Page 2 of 2 cc (with enclosure):
U.S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only)11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair -North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 BSEP 15-0022 Enclosure Brunswick Unit 2, Cycle 22 Core Operating Limits Report, March 2015 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2B21-2020 Page 1, Revision 0 BRUNSWICK UNIT 2, CYCLE 22 CORE OPERATING LIMITS REPORT March 2015 Prepared By: Verified By: Approved By: 3 /5 Ryin e. Wells BWR Fuel Engineering Peter M. Noel BWR Fuel Engineering E"WR Fuel Engineering
-Manager Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2B21-2020 Page 2, Revision 0 LIST OF EFFECTIVE PAGES PaQe(s)1-39 Revision 0 This document consists of 39 total pages.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2621-2020 B2C22 Core Operating Limits Report Page 3, Revision 0 TABLE OF CONTENTS Subject Paqe Cover ...............................................................
1 List of Effective Pages ......................................................................................................................
2 Table of Contents .............................................................................................................................
3 L is t o f T a b le s ....................................................................................................................................
4 L ist o f F ig u re s ...................................................................................................................................
5 N o m e n c la tu re ...................................................................................................................................
6 Introduction and Sum mary ........................................................................................................
8 A P L H G R L im its ................................................................................................................................
9 M C P R L im its ....................................................................................................................................
9 L H G R L im its ...................................................................................................................................
1 0 PBDA Setpoints
..............................................................................................................................
10 R B M S e tp o in ts ................................................................................................................................
1 1 Equipment Out-of-Service
..............................................................................................................
11 Single Loop Operation
....................................................................................................................
12 Inoperable Main Turbine Bypass System ..........................................
..........................................
12 Feedwater Tem perature Reduction
............................................................................................
12 R e fe re n c e s .....................................................................................................................................
1 4 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 4, Revision 0 CAUTION References to COLR Figures or Tables should be made using titles only; Figure and Table numbers may change from cycle to cycle.LIST OF TABLES Table Title Paee Table 1: R BM System Setpoints
............................................................................................
16 Table 2: RBM O perability Requirem ents ................................................................................
17 T able 3: P B D A S etpoints ....................................................................................................
..18 Table 4: Exposure Basis for Brunswick Unit 2 Cycle 22 Transient Analysis ...........................
19 Table 5: Power-Dependent MCPRp Lim its .............................................................................
20 NSS Insertion Times -BOC to < NEOC Table 6: Power-Dependent MCPRp Lim its .............................................................................
21 TSSS Insertion Times -BOC to < NEOC Table 7: Power-Dependent MCPRp Lim its .............................................................................
22 NSS Insertion Times -BOC to < EOCLB Table 8: Power-Dependent MCPRp Lim its .............................................................................
23 TSSS Insertion Times -BOC to < EOCLB Table 9: Power-Dependent MCPRp Lim its .............................................................................
24 NSS Insertion Times -BOC to < MCE (FFTR/Coastdown)
Table 10: Power-Dependent MCPRp Lim its .............................................................................
25 TSSS Insertion Times -BOC to < MCE (FFTR/Coastdown)
Table 11: Flow-Dependent M CPRf Lim its .................................................................................
26 Table 12: AREVA Fuel Steady-State LHGRss Limits ...............................................................
27 Table 13: AREVA Fuel Power-Dependent LHGRFACp Multipliers
.............................................
28 NSS Insertion Times -BOC to < EOCLB Table 14: AREVA Fuel Power-Dependent LHGRFACp Multipliers
.............................................
29 TSSS Insertion Times -BOC to < EOCLB Table 15: AREVA Fuel Power-Dependent LHGRFACp Multipliers
.............................................
30 NSS Insertion Times -BOC to < MCE (FFTR/Coastdown)
Table 16: AREVA Fuel Power-Dependent LHGRFACp Multipliers
.............................................
31 TSSS Insertion Times -BOC to < MCE (FFTR/Coastdown)
Table 17: AREVA Fuel Flow-Dependent LHGRFACf Multipliers
...............................................
32 Table 18: AREVA Fuel Steady-State MAPLHGRss Limits ........................................................
33 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Caic. No. 2B21-2020 B2C22 Core Operating Limits Report Page 5, Revision 0 I 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 O ption III Power/Flow Map ..........................................................................
34 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability O ption III Power/Flow Map ..........................................................................
35 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability O ption III Power/Flow Map ..........................................................................
36 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability O ption III Power/Flow M ap ..........................................................................
37 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability O ption III Power/Flow M ap ..........................................................................
38 OPRM Operable, FWTR, 2923 MWt Figure 6: Stability O ption III Power/Flow M ap ..........................................................................
39 OPRM Inoperable, FWTR, 2923 MWt Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design CaIc. No. 2821-2020 Page 6, Revision 0 2PT APLHGR APRM ARTS BOC BSP BWROG CAVEX COLR CRWE DIVOM EFPD EOC EOCLB EOFP EOOS F FHOOS FFTR FWTR GE HCOM HPSP HTSP ICF IPSP ITSP NOMENCLATURE Two Recirculation Pump Trip Average Planar Linear Heat Generation Rate Average Power Range Monitor (Subsystem)
APRM/RBM Technical Specification Beginning of Cycle Backup Stability Protection BWR Owners Group Core Average Exposure Core Operating Limits Report Control Rod Withdrawal Error Delta CPR Over Initial MCPR Versus Oscillation Magnitude Effective Full Power Day End of Cycle End of Cycle Licensing Basis End of Full Power Equipment Out of Service Flow (Total Core)Feedwater Heater Out of Service Final Feedwater Temperature Reduction Feedwater Temperature Reduction General Electric Hot Channel Oscillation Magnitude High Power Set Point High Trip Set Point Increased Core Flow Intermediate Power Set Point Intermediate Trip Set Point Limiting Condition of Operation Linear Heat Generation Rate Steady-State Maximum Linear Heat Generation Rate Linear Heat Generation Rate Factor Flow-Dependent Linear Heat Generation Rate Factor Power-Dependent Linear Heat Generation Rate Factor Local Power Range Monitor (Subsystem)
Low Power Set Point Lead Test Assembly Low Trip Set Point Maximum Average Planar Linear Heat Generation Rate Steady-State Maximum Average Planar Linear Heat Generation Rate Maximum Average Planar Linear Heat Generation Rate Factor LCO LHGR LHGRss LHGRFAC LHGRFACf LHGRFACp LPRM LPSP LTA LTSP MAPLHGR MAPLHGRss MAPFAC Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2B21-2020 Page 7, Revision 0 NOMENCLATURE (continued)
MAPFACf MAPFACP MAPFACSLO MCE MCPR MCPRf MCPRp MELLL MEOD MSIVOOS NEOC NFWT NRC NSS OLMCPR OPRM OOS P PBDA PRNM RBM RFWT RPT RTP SLMCPR SLO SRV SRVOOS SS STP TBV TBVINS TBVOOS TIP TLO TS TSSS Flow-Dependent Maximum Average Planar Linear Heat Generation Rate Factor Power-Dependent Maximum Average Planar Linear Heat Generation Rate Factor Maximum Average Planar Linear Heat Generation Rate Factor when in SLO Maximum Core Exposure Minimum Critical Power Ratio Flow-Dependent Minimum Critical Power Ratio Power-Dependent Minimum Critical Power Ratio Maximum Extended Load Line Limit Maximum Extended Operating Domain Main Steam Isolation Valve Out of Service Near End of Cycle Nominal Feedwater Temperature Nuclear Regulatory Commission Nominal SCRAM Speed Operating Limit Minimum Critical Power Ratio Oscillation Power Range Monitor Out of Service Power (Total Core Thermal)Period Based Detection Algorithm Power Range Neutron Monitoring (System)Rod Block Monitor (Subsystem)
Reduced Feedwater Temperature Recirculation Pump Trip Rated Thermal Power Safety Limit Minimum Critical Power Ratio Single Loop Operation Safety Relief Valve Safety Relief Valve Out of Service Steady-State Simulated Thermal Power Turbine Bypass Valve Turbine Bypass Valves In Service Turbine Bypass Valves Out of Service (all bypass valves OOS)Traversing Incore Probe Two Loop Operation Technical Specification Technical Specification SCRAM Speed Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2B21-2020 Page 8, Revision 0 CAUTION References to COLR Figures or Tables should be made using titles only; Figure and Table numbers may change from cycle to cycle.I Introduction and Summary The Brunswick Unit 2, Cycle 22 COLR provides values for the core operation limits and setpoints required by Technical Specifications (TS) 5.6.5.a.Required Core NRC Operating Limit Approved Related TS Items TS 5.6.5.a) Methodology (TS 5.6.5.b)1. APLHGR forTS 3.2.1. 1, 2, 6, 7,16, -TS 3.2.1 LCO (APLHGR)17 -TS 3.4.1 LCO (Recirculation loops operating)
-TS 3.7.6 LCO (Main Turbine Bypass out of service)2. MCPR for TS 3.2.2. 1, 2, 6, 7, 8, 9, -TS 3.2.2 LCO (MCPR)10, 11, 12, 13, -TS 3.4.1 LCO (Recirculation loops 14, 21 operating)
-TS 3.7.6 LCO (Main Turbine bypass out of service)3. LHGR for TS 3.2.3. 2, 3, 4, 5, 6, 7, -TS 3.2.3 LCO (LHGR)8, 9, 10, 12 -TS 3.4.1 LCO (Recirculation loops 13, 20 operating)
-TS 3.7.6 LCO (Main Turbine bypass out of service)4. PBDA setpointfor 8, 14, 18, 19, -TS Table 3.3.1.1-1, Function 2.f Function 2.f, APRM -OPRM 21 (APRM -OPRM Upscale)Upscale, for TS 3.3. 1.1. -TS 3.3.1.1, Condition I (Alternate instability detection and suppression)
: 5. The Allowable Values and 6, 8 -TS Table 3.3.2.1-1, Function 1 (RBM power range setpoints for Rod upscale and operability requirements)
Block Monitor Upscale Functions for TS 3.3.2.1.The required core operating limits and setpoints listed in TS 5.6.5.a are presented in the COLR, have been determined using NRC approved methodologies (COLR References 1 through 21) in accordance with TS 5.6.5.b, have considered all fuel types utilized in B2C22, and are established such that all applicable limits of the plant safety analysis are met in accordance with TS 5.6.5.c.In addition to the TS required core operating limits and setpoints, this COLR also includes maps showing the allowable power/flow operating range including the Option III stability ranges.The generation of this COLR is documented in Reference 30 and is based on analysis results documented in References 27-29.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 9, Revision 0 APLHGR Limits Steady-state MAPLHGRss limits are provided for AREVA Fuel (Table 18). These steady-state MAPLHGRss limits must be modified as follows:* AREVA Fuel MAPLHGR limits do not have a power, flow, or EOOS dependency.
Power-dependent MAPFACP multipliers and flow-dependent MAPFACf multipliers with a constant value of 1.0 under all conditions have been assigned to AREVA Fuel.* 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 x (MAPFACp, MAPFACf, 1.0)min MAPLHGR LimitSLO = MAPLHGRss x (MAPFACp, MAPFACf, MAPFACSLO)min 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 11 are based on the TLO and SLO SLMCPRs listed in Technical Specification 2.1.1.2 as >1.08 and >1.11, 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 +0.03 when in SLO.The applied TLO and SLO MCPR limits are determined as follows: MCPR LimitTLO = (MCPRp, MCPRf)max MCPR LimitSLO = (MCPRp + 0.03, MCPRf)max Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show two breakpoints at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0), THEN select the most restrictive limit associated with the breakpoint.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 10, Revision 0 LHGR Limits Steady-state LHGRss limits are provided for AREVA Fuel (Table 12). These steady-state LHGRss limits must be modified as follows:* AREVA Fuel LHGR limits have a core power and core flow dependency.
AREVA 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." AREVA 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 two breakpoints at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0), THEN select the most restrictive limit associated with the breakpoint.
PBDA Setpoints Brunswick Unit 2 has implemented BWROG Long Term Stability Solution Option III (OPRM) with the methodology described in Reference
: 23. Plant specific analysis incorporating the Option III hardware is described in Reference
: 24. Reload validation has been performed in accordance with Reference 19.The analysis was performed at 100%P assuming a two pump trip (2PT) and at 45%F assuming steady-state (SS) conditions at the highest rod line power (60.5%). The PBDA setpoints are set such that either the least limiting MCPRp limit or the least limiting MCPRf limit will provide adequate protection against violation of the SLMCPR during a postulated reactor instability.
Based on the MCPR limits presented in Tables 5 through 11, the required Amplitude Trip Setpoint (1.10) is set by the least limiting 100%P MCPRp limit (1.34) with an allowance for conservative margin, which has an associated Confirmation Count Setpoint (13). The PBDA setpoints shown in Table 3 are valid for any feedwater temperature.
Evaluations by GE have shown that the generic DIVOM curves specified in Reference 19 may not be conservative for current plant operating conditions for plants which have implemented Stability Option Ill. To address this issue, AREVA has performed calculations for the relative change in CPR as a function of the calculated HCOM. These calculations were performed with the RAMONA5-FA code in accordance with Reference
: 26. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in ACPR and the HCOM on a plant specific basis. The stability-based OLMCPRs are based upon using the most limiting ACPR calculated for a given oscillation magnitude or the generic value provided in Reference 19.In cases where the OPRM system is declared inoperable, Backup Stability Protection (BSP) in accordance with Reference 25 is provided.
Analyses have been performed to support operation with nominal feedwater temperature conditions and reduced feedwater temperature conditions Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 11, Revision 0 (FHOOS and FFTR).The power/flow maps (Figures 1-6) were validated for B2C22 based on Reference 29 to facilitate operation under Stability Option III as implemented by Function 2.f of Table 3.3.1.1-1 and LCO Condition I of Technical Specification 3.3.1.1. The generation of these maps is documented in Reference
: 28. All maps illustrate the region of the power/flow map above 25% RTP and below 60%drive flow (correlated to core flow) where the system is required to be enabled. Figures 1-6 were included in the COLR as an operator aid and not a licensing requirement.
Figures 5 and 6 are the power/flow maps for use in FWTR.The maps supporting an operable OPRM (Figures 1, 3 and 5) show a Scram Avoidance Region, which is not a licensing requirement but is an operator aid to illustrate where the OPRM system may generate a scram to avoid an instability event. Note that the STP scram and rod block limits are defined in Technical Specifications, the Technical Requirements Manual, and/or Plant procedures, and are included in the COLR as an operator aid rather than a licensing requirement.
Figures 3 and 4 implement the corrective action for AR-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 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.
0 Any 1 inoperable SRV* 2 inoperable TBV (Note that for TBVOOS and TBVOOS/FHOOS all 10 TBVs are assumed inoperable)
* Up to 40% of the TIP channels OOS 0 Up to 50% of the LPRMs OOS Please note that during FFTR/Coastdown, FHOOS is included in Base Case Operation and TBVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design CaIc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 12, Revision 0 Single Loop Operation Brunswick Unit 2, 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. 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.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.Inoperable Main Turbine Bypass System Brunswick Unit 2, Cycle 22 may operate with an inoperable Main Turbine Bypass System over the entire MEOD range and cycle with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. An operable Main Turbine Bypass System with only two inoperable bypass valves was assumed in the development of the Base Case Operation limits. Base Case Operation is synonymous with TBVINS. The following must be considered when operating with TBVOOS: " Three or more inoperable bypass valves renders the entire Main Turbine Bypass System inoperable requiring the use of TBVOOS limits. The TBVOOS analysis supports operation with all bypass valves inoperable.
* Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >1O&deg;F and reactor power > 50% RTP requires use of the TBVOOS/FHOOS limits. At or below 50% RTP, TBVOOS limits bound FHOOS limits.* TBVOOS operation coincident with FHOOS is supported using the combined TBVOOS/FHOOS limits." SLO is not permitted with TBVOOS.Feedwater Temperature Reduction Brunswick Unit 2, Cycle 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 -10'F. NFWT and its allowable variation were assumed in the development of the Base Case Operation limits. The FHOOS limits and FFTR/Coastdown limits were developed for a maximum feedwater temperature reduction of 110.3&deg;F.The following must be considered when operating with RFWT:* Although the acronyms FWTR, FHOOS, RFWT and FFTR all involve reduced feedwater temperature, the use of FFTR is reserved for cycle energy extension using reduced feedwater Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design CaIc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 13, Revision 0 temperature at and beyond a core average exposure of EOCLB using FFTR/Coastdown limits.* Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >1O&deg;F and reactor power > 50% RTP requires use of the FHOOS limits. At or below 50% RTP, Base Case Operation limits bound FHOOS limits." Until a core average exposure of EOCLB is reached, implementation of the FFTR/Coastdown limits is not required even if coastdown begins early." When operating with RFWT, the appropriate Stability Option III Power/Flow Maps (Figures 5 and 6) must be used." FHOOS operation coincident with TBVOOS is supported using the combined TBVOOS/FHOOS limits." SLO is not permitted with RFWT.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design CaIc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 14, Revision 0 References In accordance with Brunswick Unit 2 Technical Specification 5.6.5.b, the analytical methods for determining Brunswick Unit 2 core operating limits have been specifically reviewed and approved by the NRC and are listed as References 1 through 21.1. NEDE-2401 1-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 Supplement 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis," March 1983.7. XN-NF-80-19(P)(A)
Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Revision 1, June 1986.8. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Revision 0, October 1999.9. XN-NF-80-19(P)(A)
Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Revision 2, January 1987.10. XN-NF-84-105(P)(A)
Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T:
A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987.11. ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," Revision 0, June 2011.12. ANF-913(P)(A)
Volume 1 and Volume 1 Supplements 2, 3, 4, "COTRANSA2:
A Computer Program for Boiling Water Reactor Transient Analyses," Revision 1, August 1990.13. ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Revision 3, September 2005.14. EMF-2209(P)(A), "SPCB Critical Power Correlation," Revision 3, September 2009.15. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Revision 0, August 2000.16. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model," Revision 0, May 2001.17. EMF-2292(P)(A), "ATRIUMTM-10:
Appendix K Spray Heat Transfer Coefficients," Revision 0, September 2000.18. EMF-CC-074(P)(A)
Volume 4, "BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2," Revision 0, August 2000.19. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.20. BAW-1 0247PA, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," Revision 0, April 2008.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 15, Revision 0 21. ANP-10298PA, "ACE/ATRIUM 10XM Critical Power Correlations," Revision 0, March 2010.22. NEDC-31654P, "Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant," February 1989.23. NEDO-31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology (Supplement 1)," November 1995, NRC ADAMS Accession No. ML14093A211.
: 24. GENE-C51-00251-00-01, "Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick 1 and 2," Revision 0, March 2001.25. OG 02-0119-260 "Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy," July 17, 2002.26. BAW-10255PA, "Cycle Specific DIVOM Methodology Using the RAMONA5-FA Code," Revision 2, May 2008.27. BNP Design Calculation 2C51-0001, "Power Range Neutron Monitoring System Setpoint Uncertainty and Scaling Calculation (2-C51-APRM-1 through 4 Loops and 2-C51-RBM-A and B Loops)," Revision 3, May 2004.28. BNP Design Calculation 0B21-1015, "BNP Power/Flow Maps," Revision 7, March 2008.29. ANP-3369(P), "Brunswick Unit 2 Cycle 22 Reload Safety Analysis," Revision 0, January 2015.30. BNP Design Calculation 2B21-2020, "Preparation of the B2C22 Core Operating Limits Report," Revision 0.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2821-2020 Page 16, Revision 0 Table 1 RBM System Setpoints 1 Setpoint a Setpoint Value Allowable Value Lower Power Setpoint (LPSP b) < 27.7 < 29.0 Intermediate Power Setpoint (IPSPb) < 62.7 < 64.0 High Power Setpoint (HPSP b) < 82.7 < 84.0 Low Trip Setpoint (LTSPc' d) < 114.1 < 114.6 Intermediate Trip Setpoint (ITSPc'd)
< 108.3 < 108.8 High Trip Setpoint (HTSPc'd)
< 104.5 < 105.0 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, < 114.1 means< 114.1/125.0 of full scale.d Trip setpoints and allowable values are based on a HTSP Analytical Limit of 107.4 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 82C22 Core Operating Limits Report Design Calc. No. 2B21-2020 Page 17, Revision 0 Table 2 RBM Operability Requirements 2 IF the following conditions are met, THEN RBM Not Required Operable Thermal Power ATRIUM 1OXM ATRIUM 11 LTA (% rated) MCPR MCPR 2 <1.91 TLO -1.55 TLO 291.96 SLO ->1.60 SLO-90% -1.52 TLO ->1.36 TLO 2 Requirements valid for all fuel designs, all SCRAM insertion times and all core average exposure ranges.
Duke Energy, Nuclear Fuels Engineering, B2C22 Core Operating Limits Report Nuclear Fuel Design Table 3 PBDA Setpoints 3 Design Calc. No. 2B21-2020 Page 18, Revision 0 Amplitude Trip OLMCPR(SS)
OLMCPR(2PT)
Setpoint (SP)1.05 1.20 1.21 1.06 1.22 1.23 1.07 1.23 1.25 1.08 1.25 1.27 1.09 1.27 1.29 1.10 1.29 1.31 1.11 1.31 1.33 1.12 1.33 1.35 1.13 1.35 1.37 1.14 1.37 1.39 1.15 1.40 1.41 Acceptance Criteria Off-rated OLMCPR @ Rated Power 45% Flow OLMCPR 3 This table is referred to by Technical Specification 3.3.1.1 (Table 3.3.1.1-1) and 5.6.5.a.4.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Table 4 Exposure Basis 4 for Brunswick Unit 2 Cycle 22 Transient Analysis Design Calc. No. 2B21-2020 Page 19, Revision 0 Core Average Exposure Comments (MWd/MTU)33,206 Break point for exposure-dependent MCPR, limits (NEOC)35,012 Design basis rod patterns to EOFP + 15 EFPD (EOCLB)36,797 End of cycle with FFTR/Coastdown
-Maximum Core Exposure (MCE)4 The exposure basis for the defined break points is the core average exposure (CAVEX) values shown above regardless of the actual BOC CAVEX value of the As-Loaded Core.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design CaIc. No. 2B21-2020 Page 20, Revision 0 Table 5 Power-Dependent MCPRp Limits 5 NSS Insertion Times BOC to < NEOC EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) MCPRp MCPRp 100.0 1.34 1.35 90.0 1.39 1.37 50.0 1.67 1.53 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.37 1.37 90.0 1.40 1.40 50.0 1.67 1.54> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 100.0 1.34 1.35 90.0 1.39 1.37 50.0 1.67 1.53> 65%F 5 65%F > 65%F < 65%F 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.37 1.37 90.0 1.40 1.40 50.0 1.67 1.57 TBVOOS > 65%F < 65%F > 65%F < 65%F and and 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 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 0.03. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Table 6 Power-Dependent MCPRp Limits 6 TSSS Insertion Times BOCto<NEOC Design Calc. No. 2B21-2020 Page 21, Revision 0 EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) MCPRp MCPRp 100.0 1.38 1.38 90.0 1.39 1.40 50.0 1.68 1.53 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.40 1.41 90.0 1.43 1.43 50.0 1.68 1.56> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 100.0 1.38 1.38 90.0 1.39 1.40 50.0 1.68 1.53> 65%F < 65%F > 65%F < 65%F 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.40 1.41 90.0 1.43 1.43 TBVOOS 50.0 1.68 1.60 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.95 2.65 3.13 2.85 1 23.0 3.14 2.85 3.31 3.08 6 Limits support operation with any combination of any 1 inoperable SRV, 2 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.
For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.03. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Caic. No. 2B21-2020 Page 22, Revision 0 Table 7 Power-Dependent MCPRp Limits 7 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) MCPRp MCPRp 100.0 1.35 1.37 90.0 1.39 1.39 50.0 1.67 1.53 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.37 1.39 90.0 1.40 1.42 50.0 1.67 1.54> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 100.0 1.35 1.37 90.0 1.39 1.39 50.0 1.67 1.53> 65%F < 65%F > 65%F < 65%F 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.37 1.39 90.0 1.40 1.42 TBVOOS 50.0 1.67 1.57 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 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 0.03. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Table 8 Power-Dependent MCPRp Limits 8 TSSS Insertion Times BOC to < EOCLB Design CaIc. No. 2B21-2020 Page 23, Revision 0 EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) MCPRp MCPRp 100.0 1.40 1.41 90.0 1.40 1.42 50.0 1.68 1.53 Base Case > 65%F < 65%F > 65%F s 65%F Operation 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.42 1.44 90.0 1.44 1.46 50.0 1.68 1.56> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 100.0 1.40 1.41 90.0 1.40 1.42 50.0 1.68 1.53> 65%F 5 65%F > 65%F < 65%F 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.20 2.05 2.27 2.11 23.0 2.26 2.11 2.33 2.18 100.0 1.42 1.44 90.0 1.44 1.46 TBVOOS 50.0 1.68 1.60 and > 65%F < 65%F > 65%F < 65%F FHOOS 50.0 1.88 1.73 1.91 1.75 26.0 2.19 2.04 2.26 2.09 26.0 2.95 2.65 3.13 2.85 1 23.0 3.14 2.85 3.31 3.08 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 0.03. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Table 9 Power-Dependent MCPRP Limits 9 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)
Design CaIc. No. 2B21-2020 Page 24, Revision 0 EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) MCPR, MCPRP Base Case 100.0 1.36 1.38 Operation 90.0 1.39 1.40 50.0 1.67 1.53 (FFTR/FHOOS
> 65%F < 65%F > 65%F  65%F included) 50.0 1.86 1.71 1.89 1.74 26.0 2.18 2.04 2.25 2.08 (Bounds operation 26.0 2.20 2.05 2.27 2.11 with NFWT) 23.0 2.26 2.11 2.33 2.18 100.0 1.37 1.39 TBVOOS 90.0 1.40 1.42 50.0 1.67 1.57 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 1.86 1.71 1.89 1.74 (Bounds operation 26.0 2.18 2.04 2.25 2.08 with NFWT) 26.0 2.95 2.65 3.13 2.85 23.0 3.14 2.85 3.31 3.08 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 0.03. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2B21-2020 Page 25, Revision 0 Table 10 Power-Dependent MCPRP Limits 1&deg;TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) MCPRP MCPRp Base Case 100.0 1.44 1.48 Operation 90.0 1.45 1.51 50.0 1.68 1.55 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 1.88 1.73 1.93 1.77 26.0 2.19 2.04 2.28 2.11 (Bounds operation 26.0 2.20 2.05 2.29 2.13 with NFWT) 23.0 2.26 2.11 2.35 2.20 100.0 1.46 1.50 TBVOOS 90.0 1.46 1.51 50.0 1.70 1.64 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 1.90 1.75 1.94 1.78 (Bounds operation 26.0 2.21 2.06 2.29 2.12 with NFWT) 26.0 2.97 2.67 3.16 2.88 23.0 3.16 2.87 3.34 3.11 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 0.03. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Caic. No. 2B21-2020 Page 26, Revision 0 Table 11 Flow-Dependent MCPRf Limits 1 1 Core Flow ATRIUM 1OXM ATRIUM 11 LTA (% of rated) MCPRf MCPRf 0.0 1.70 1.70 31.0 1.70 1.70 55.0 1.59 1.59 100.0 1.20 1.20 107.0 1.20 1.20 11 Limits valid for all SCRAM insertion times and all core average exposure ranges.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Caic. No. 2B21-2020 Page 27, Revision 0 Table 12 AREVA Fuel Steady-State LHGRss Limits Peak ATRIUM 1OXM ATRIUM 11 LTA Pellet Exposure LHGR LHGR (GWd/MTU) (kW/ft) (kW/ft)0.0 14.1 12.2 18.9 14.1 12.2 74.4 7.4 6.4 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design CaIc. No. 2B21-2020 Page 28, Revision 0 Table 13 AREVA Fuel Power-Dependent LHGRFACP Multipliers 1 2 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) LHGRFACp LHGRFACP 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.39 0.46 0.39 0.46 23.0 0.36 0.42 0.36 0.42 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92> 65%F < 65%F > 65%F < 65%F 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 TBVOOS> 65%F < 65%F > 65%F < 65%F and Fnd 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.39 0.46 0.39 0.46 23.0 0.36 0.42 0.36 0.42 12 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.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design CaIc. No. 2B21-2020 B2C22 Core Operating Limits Report Page 29, Revision 0 Table 14 AREVA Fuel Power-Dependent LHGRFACP Multipliers 1 3 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) LHGRFAC, LHGRFAC, 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 Base Case > 65%F < 65%F > 65%F < 65%F Operation 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92> 65%F < 65%F > 65%F < 65%F TBVOOS 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.39 0.46 0.39 0.46 23.0 0.36 0.42 0.36 0.42 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92> 65%F < 65%F > 65%F < 65%F 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 26.0 0.64 0.66 0.64 0.66 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 90.0 1.00 1.00 50.0 0.92 0.92 TBVOOS and > 65%F < 65%F > 65%F < 65%F FHd 50.0 0.86 0.86 0.86 0.86 FHOOS 26.0 0.64 0.66 0.64 0.66 26.0 0.39 0.46 0.39 0.46 23.0 0.36 0.42 0.36 0.42 13 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.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design CaIc. No. 2B21-2020 Page 30, Revision 0 Table 15 AREVA Fuel Power-Dependent LHGRFACP Multipliers 1 4 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) LHGRFACp LHGRFACp Base Case 100.0 1.00 1.00 Operation 90.0 1.00 1.00 50.0 0.92 0.92 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 TBVOOS 90.0 1.00 1.00 50.0 0.92 0.92 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 0.86 0.86 0.86 0.86 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 26.0 0.39 0.46 0.39 0.46 23.0 0.36 0.42 0.36 0.42 14 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.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design CaIc. No. 2B21-2020 Page 31, Revision 0 Table 16 AREVA Fuel Power-Dependent LHGRFACP Multipliers 1 5 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 1OXM ATRIUM 11 LTA Condition
(% rated) LHGRFACP LHGRFACP Base Case 100.0 1.00 1.00 Operation 90.0 1.00 1.00 50.0 0.92 0.92 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 0.86 0.86 0.86 0.86 26.0 0.64 0.66 0.64 0.66 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 23.0 0.60 0.64 0.60 0.64 100.0 1.00 1.00 TBVOOS 90.0 1.00 1.00 50.0 0.92 0.92 (FFTR/FHOOS
> 65%F < 65%F > 65%F < 65%F included) 50.0 0.86 0.86 0.86 0.86 (Bounds operation 26.0 0.64 0.66 0.64 0.66 with NFWT) 26.0 0.39 0.46 0.39 0.46 23.0 0.36 0.42 0.36 0.42 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.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design Calc. No. 2621-2020 Page 32, Revision 0 Table 17 AREVA Fuel Flow-Dependent LHGRFACf Multipliers 1 6 Core Flow ATRIUM 10XM ATRIUM 11 LTA (% of rated) LHGRFACf LHGRFACf 0.0 0.58 0.58 31.0 0.58 0.58 75.0 1.00 1.00 107.0 1.00 1.00 16 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Design CaIc. No. 2B21-2020 Page 33, Revision 0 Table 18 AREVA Fuel Steady-State MAPLHGRss Limits 1 7' 18 Average Planar Exposure ATRIUM 1OXM ATRIUM 11 LTA (GWd/MTU)
MAPLHGR MAPLHGR (kW/ft) (kW/ft)0.0 13.1 10.5 15.0 13.1 10.5 67.0 7.7 5.9 17 AREVA Fuel MAPLHGR limits do not have a power or flow dependency.
Thus, the ATRIUM 1OXM and ATRIUM 11 MAPFACp and the MAPFACf multipliers have a constant value of 1.0 under all conditions.
18 ATRIUM 1OXM 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 B2C22 Core Operating Limits Report Figure 1 Stability Option III Power/Flow Map OPRM Operable, Two Loop Operation, 2923 MWt Design Calc. No. 2B21-2020 Page 34, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3:]120.0 110.0 100.0 90.0 80.0 70.0 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow i Mlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbslhr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 %Core Flow
 
==Reference:==
 
0B21-1015, Revision 7 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Figure 2 Stability Option III Power/Flow Map OPRM Inoperable, Two Loop Operation, 2923 MWt Design CaIc. No. 2B21-2020 Page 35, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow% Mlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
 
==Reference:==
 
0B21-1015, Revision 7 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Figure 3 Stability Option III Power/Flow Map OPRM Operable, Single Loop Operation, 2923 MWt Design Calc. No. 2B21-2020 Page 36, Revision 0 This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow Mlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
 
==Reference:==
 
0B21-1015, Revision 7 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Figure 4 Stability Option III Power/Flow Map OPRM Inoperable, Single Loop Operation, 2923 MWt Design Calc. No. 2B21-2020 Page 37, Revision 0 This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MEULL) (ICF)Core Core Power Flow Flow%e M b___hr_ Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
 
==Reference:==
 
0B21-1015, Revision 7 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Figure 5 Stability Option III Power/Flow Map OPRM Operable, FWTR, 2923 MWt Design CaIc. No. 2B21-2020 Page 38, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 1 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow FlowMlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
 
==Reference:==
 
0B21-1015, Revision 7 Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design B2C22 Core Operating Limits Report Figure 6 Stability Option III Power/Flow Map OPRM Inoperable, FWTR, 2923 MWt Design Calc. No. 2B21-2020 Page 39, Revision 0 I This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 110.0 100.0 90.0 80.0 70.0 o 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum Maximum (MELLL) (ICF)Core Core Power Flow Flow% Mlbs/hr Mlbs/hr 100 76.19 80.47 99 75.04 80.47 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 86 60.58 80.60 85 59.50 80.69 84 58.43 80.79 83 57.37 80.90 82 56.31 81.05 81 55.25 81.21 80 54.20 81.36 79 53.16 81.51 78 52.12 81.67 77 51.08 81.82 76 50.05 81.98 75 49.02 82.13 74 48.00 82.29 73 46.98 82.44 72 45.96 82.60 71 44.95 82.75 70 43.94 82.91 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 62 36.04 84.14 61 35.06 84.30 60 34.10 84.45 59 33.13 84.61 58 32.17 84.70 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
 
==Reference:==
 
0B21-11015, Revision 7}}

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