BSEP 18-0042, Cycle 22 Core Operating Limits Report (COLR)
| ML18086B639 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/27/2018 |
| From: | Wooten B Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 18-0042 | |
| Download: ML18086B639 (42) | |
Text
( -, DUKE Brunswick Nuclear Plant P.O. Box 10429 ENERGYe Southport, NC 28461 March 27, 2018 Serial: BSEP 18-0042 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Brunswick Steam Electric Plant, Unit No. 1 Renewed Facility Operating License No. DPR-71 Docket No. 50-325 Unit 1 Cycle 22 Core Operating Limits Report (COLR)
Reference Letter from Annette H. Pope (Duke Energy) to NRC Document Control Desk, Unit 1 Cycle 21 Core Operating Limits Report (COLR), dated March 18, 2016 Ladies and Gentlemen:
Enclosed is a copy of the Core Operating Limits Report (COLR) for Brunswick Steam Electric Plant (BSEP), Unit 1 Cycle 22 operation. Duke Energy Progress, LLC (Duke Energy), is providing the enclosed COLR in accordance with Brunswick Unit 1 Technical Specification 5.6.5.d. The enclosed COLR supersedes the report previously submitted by letter dated March 18, 2016 (i.e., Reference).
This letter and the enclosed COLR do not contain any regulatory commitments.
Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 832-2487.
Sincerely, 0-~-~.:i?
Bryan B. Wooten Director - Organizational Effectiveness Brunswick Steam Electric Plant MAT/mat
Enclosure:
Brunswick Unit 1, Cycle 22 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. Andrew Hon (Mail Stop OWFN 8G9A) 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
BSEP 18-0042 Enclosure Brunswick Unit 1, Cycle 22 Core Operating Limits Report
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Cale. No. 1921-2050 Rev 0 B1C22 Core Operating Limits Report, BNEl*0400-0017 Rev. 0 Page 1 BRUNSWICK UNIT 1, CYCLE 22 CORE OPERATING LIMITS REPORT March 2018 Prepared by:
teshA. Kuntawala Brunswick Nuclear Design Reviewed by:
Site Inspection by:
T. Brya eater Bruna ck Reactor Engineering Approved by:
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 2 LIST OF EFFECTIVE PAGES Page(s) Revision 1- 39 0 This document consists of 39 total pages.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 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 PBDA Setpoints ............................................................................................................................. 10 RBM Setpoints ............................................................................................................................... 11 Equipment Out-of-Service .............................................................................................................. 11 Single Loop Operation.................................................................................................................... 12 Inoperable Main Turbine Bypass System ....................................................................................... 12 Feedwater Temperature Reduction ................................................................................................ 13 References..................................................................................................................................... 14
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 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: PBDA Setpoints ............................................................................................................ 18 Table 4: Exposure Basis for Brunswick Unit 1 Cycle 22 Transient Analysis ................................ 19 Table 5: Power-Dependent MCPRp Limits .................................................................................. 20 NSS Insertion Times - BOC to < NEOC Table 6: Power-Dependent MCPRp Limits .................................................................................. 21 TSSS Insertion Times - BOC to < NEOC Table 7: Power-Dependent MCPRp Limits .................................................................................. 22 NSS Insertion Times - BOC to < EOCLB Table 8: Power-Dependent MCPRp Limits .................................................................................. 23 TSSS Insertion Times - BOC to < EOCLB Table 9: Power-Dependent MCPRp Limits .................................................................................. 24 NSS Insertion Times - BOC to < MCE (FFTR/Coastdown)
Table 10: Power-Dependent MCPRp Limits .................................................................................. 25 TSSS Insertion Times - BOC to < MCE (FFTR/Coastdown)
Table 11: Flow-Dependent MCPRf Limits ..................................................................................... 26 Table 12: 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 Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 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 Option III Power/Flow Map .............................................................................. 34 OPRM Operable, Two Loop Operation, 2923 MWt Figure 2: Stability Option III Power/Flow Map .............................................................................. 35 OPRM Inoperable, Two Loop Operation, 2923 MWt Figure 3: Stability Option III Power/Flow Map .............................................................................. 36 OPRM Operable, Single Loop Operation, 2923 MWt Figure 4: Stability Option III Power/Flow Map .............................................................................. 37 OPRM Inoperable, Single Loop Operation, 2923 MWt Figure 5: Stability Option III Power/Flow Map .............................................................................. 38 OPRM Operable, FWTR, 2923 MWt Figure 6: Stability Option III Power/Flow Map .............................................................................. 39 OPRM Inoperable, FWTR, 2923 MWt
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 6 NOMENCLATURE 2PT Two Recirculation Pump Trip APLHGR Average Planar Linear Heat Generation Rate APRM Average Power Range Monitor (Subsystem)
ARTS APRM/RBM Technical Specification BOC Beginning of Cycle BSP Backup Stability Protection BWROG BWR Owners Group CAVEX Core Average Exposure COLR Core Operating Limits Report CRWE Control Rod Withdrawal Error DIVOM Delta CPR Over Initial MCPR Versus Oscillation Magnitude EFPD Effective Full Power Day EOC End of Cycle EOCLB End of Cycle Licensing Basis EOFP End of Full Power EOOS Equipment Out-of-Service F Flow (Total Core)
FHOOS Feedwater Heater Out-of-Service FFTR Final Feedwater Temperature Reduction FWTR Feedwater Temperature Reduction GE General Electric HCOM Hot Channel Oscillation Magnitude HPSP High Power Set Point HTSP High Trip Set Point ICF Increased Core Flow IPSP Intermediate Power Set Point ITSP Intermediate Trip Set Point LCO Limiting Condition of Operation LHGR Linear Heat Generation Rate 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 LPRM Local Power Range Monitor (Subsystem)
LPSP Low Power Set Point LTSP Low Trip Set Point 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
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 7 NOMENCLATURE (continued)
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 MEOD Maximum Extended Operating Domain MSIVOOS Main Steam Isolation Valve Out-of-Service NEOC Near End of Cycle NFWT Nominal Feedwater Temperature NRC Nuclear Regulatory Commission NSS Nominal SCRAM Speed OLMCPR Operating Limit Minimum Critical Power Ratio OPRM Oscillation Power Range Monitor OOS Out-of-Service P Power (Total Core Thermal)
PBDA Period Based Detection Algorithm PRNM Power Range Neutron Monitoring (System)
RBM Rod Block Monitor (Subsystem)
RFWT Reduced Feedwater Temperature RPT Recirculation Pump Trip RTP Rated Thermal Power SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRV Safety Relief Valve SRVOOS Safety Relief Valve Out-of-Service SS Steady-State STP Simulated Thermal Power TBV Turbine Bypass Valve TBVINS Turbine Bypass Valves In Service TBVOOS Turbine Bypass Valves Out-of-Service (all bypass valves OOS)
TIP Traversing Incore Probe TLO Two Loop Operation TS Technical Specification TSSS Technical Specification SCRAM Speed
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 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.
Introduction and Summary The Brunswick Unit 1, Cycle 22 COLR provides values for the core operation limits and setpoints required by Technical Specifications (TS) 5.6.5.a.
NRC Required Core Approved Operating Limit Related TS Items Methodology (TS 5.6.5.a )
- 1. APLHGR for TS 3.2.1. 1, 2, 6, 7,16, TS 3.2.1 LCO (APLHGR) 17 TS 3.4.1 LCO (Recirculation loops operating)
TS 3.7.6 LCO (Main Turbine Bypass out-of-service)
- 2. MCPR for TS 3.2.2. 1, 2, 6, 7, 8, 9, TS 3.2.2 LCO (MCPR) 10, 11, 12, 13, TS 3.4.1 LCO (Recirculation loops 14, 21 operating)
TS 3.7.6 LCO (Main Turbine bypass out-of-service)
- 3. LHGR for TS 3.2.3. 2, 3, 4, 5, 6, 7, TS 3.2.3 LCO (LHGR) 8, 9, 10, 12, TS 3.4.1 LCO (Recirculation loops 13, 20 operating)
TS 3.7.6 LCO (Main Turbine bypass out-of-service)
- 4. PBDA setpoint for 8, 14, 18, 19, TS Table 3.3.1.1-1, Function 2.f Function 2.f, APRM - OPRM 21 (APRM - OPRM Upscale)
Upscale, for TS 3.3.1.1.
TS 3.3.1.1, Condition I (Alternate instability detection and suppression)
- 5. The Allowable Values and 6, 8 TS Table 3.3.2.1-1, Function 1 (RBM power range setpoints for Rod upscale and operability requirements)
Block Monitor Upscale Functions for TS 3.3.2.1.
The required core operating limits and setpoints listed in TS 5.6.5.a are presented in the COLR, have been determined using NRC approved methodologies (COLR References 1 through 21) in accordance with TS 5.6.5.b, have considered all fuel types utilized in B1C22, and are established such that all applicable limits of the plant safety analysis are met in accordance with TS 5.6.5.c.
In addition to the TS required core operating limits and setpoints, this COLR also includes maps showing the allowable power/flow operating range including the 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. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 9 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.
The applied MAPLHGR limit is dependent on the number of recirculation loops in operation. The steady-state MAPLHGR limit must be modified by a MAPFACSLO multiplier when in SLO.
MAPFACSLO has a fuel design dependency as shown below.
The applied TLO and SLO MAPLHGR limits are determined as follows:
MAPLHGR LimitTLO = MAPLHGRSS MAPLHGR LimitSLO = MAPLHGRSS x MAPFACSLO where MAPFACSLO = 0.80 for ATRIUM 10XM Linear interpolation should be used to determine intermediate values between the values listed in the table.
MCPR Limits The MCPR limits presented in Tables 5 through 11 are based on the TLO and SLO SLMCPRs listed in Technical Specification 2.1.1.2 of 1.07 and 1.09, respectively.
MCPR limits have a core power and core flow dependency. Power-dependent MCPRp limits are presented in Tables 5 through 10 while flow-dependent MCPRf limits are presented in Table 11.
Power-dependent MCPRP limits are dependent on CAVEX, SCRAM insertion speed, EOOS, fuel design, number of operating recirculation loops (i.e., TLO or SLO), core flow and core thermal power. Values for the CAVEX breakpoints are provided in Table 4. See COLR section titled Equipment Out-of-Service for a list of analyzed EOOS conditions. Care should be used when selecting the appropriate limits set.
The MCPR limits are established such that they bound all pressurization and non-pressurization events.
The power-dependent MCPRp limits (Tables 5-10) must be adjusted by an adder of +0.02 when in SLO.
The applied TLO and SLO MCPR limits are determined as follows:
MCPR LimitTLO = (MCPRp, MCPRf)max MCPR LimitSLO = (MCPRp + 0.02, MCPRf)max Linear interpolation should be used to determine intermediate values between the values listed in the tables. Some of the limits tables show step changes at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0), THEN select the most restrictive limit associated with the breakpoint.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 10 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 step changes at 26.0%P and 50.0%P. IF performing a hand calculation of a limit AND the power is exactly on the breakpoint (i.e. 26.0 or 50.0), THEN select the most restrictive limit associated with the breakpoint.
PBDA Setpoints Brunswick Unit 1 has implemented BWROG Long Term Stability Solution Option III (OPRM) with the methodology described in Reference 23. Plant specific analysis incorporating the Option III hardware is described in Reference 24. Reload validation has been performed in accordance with Reference 19. The analysis was performed at 100%P assuming a two pump trip (2PT) and at 45%F assuming steady-state (SS) conditions at the highest rod line power (60.6%). The PBDA setpoints are set such that either the least limiting 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 (1.11) is set by the least limiting 100%P MCPRp limit (1.35) with an allowance for conservative margin, which has an associated Confirmation Count Setpoint (14). The PBDA setpoints shown in Table 3 are valid for any feedwater temperature.
Evaluations by GE have shown that the generic DIVOM curves specified in Reference 19 may not be conservative for current plant operating conditions for plants which have implemented Stability Option III.
To address this issue, AREVA has performed calculations for the relative change in CPR as a function of the calculated HCOM. These calculations were performed with the RAMONA5-FA code in accordance with Reference 26. This code is a coupled neutronic-thermal-hydraulic three-dimensional transient model for the purpose of determining the relationship between the relative change in CPR and the HCOM on a plant specific basis. The stability-based OLMCPRs are based upon using the most limiting CPR calculated for a given oscillation magnitude or the generic value provided in Reference 19.
In cases where the OPRM system is declared inoperable, Backup Stability Protection (BSP) in accordance with Reference 25 is provided. Analyses have been performed to support operation with nominal feedwater temperature conditions and reduced feedwater temperature conditions (FHOOS and FFTR).
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 11 The power/flow maps (Figures 1-6) were validated for B1C22 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 1, Cycle 22 is analyzed for the following operating conditions with applicable MCPR, APLHGR and LHGR limits.
Base Case Operation SLO TBVOOS FHOOS Combined TBVOOS and FHOOS Base Case Operation as well as the above-listed EOOS conditions assume all the items OOS below.
These conditions are general analysis assumptions used to ensure conservative analysis results and were not meant to define specific EOOS conditions beyond those already defined in Technical Specifications.
Any 1 inoperable SRV 1 inoperable TBV (Note that for TBVOOS, TBVOOS/FHOOS, all 4 TBVs are assumed inoperable)
Up to 40% of the TIP channels OOS Up to 50% of the LPRMs OOS Please note that during FFTR/Coastdown, FHOOS is included in Base Case Operation, and TBVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 12 Single Loop Operation Brunswick Unit 1, Cycle 22 may operate in SLO up to a maximum core flow of 45 Mlbm/hr which corresponds to a maximum power level of 71.1% RTP with applicable MCPR, APLHGR and LHGR limits.
The following must be considered when operating in SLO:
SLO is not permitted with RFWT (FHOOS/FFTR).
SLO is not permitted with TBVOOS.
SLO is not permitted with MSIVOOS.
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 1, 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 one inoperable bypass valve was assumed in the development of the Base Case Operation limits. Base Case Operation is synonymous with TBVINS. The following must be considered when operating with TBVOOS:
Two or more inoperable bypass valves renders the entire Main Turbine Bypass System inoperable requiring the use of TBVOOS limits. The TBVOOS analysis supports operation with all bypass valves inoperable.
Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10F and reactor power 23% RTP requires use of the combined TBVOOS/FHOOS limits.
TBVOOS operation coincident with FHOOS is supported using the combined TBVOOS/FHOOS limits.
SLO is not permitted with TBVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 13 Feedwater Temperature Reduction Brunswick Unit 1, Cycle 22 may operate with RFWT over the entire MEOD range and cycle with applicable APLHGR, MCPR and LHGR limits as specified in the COLR. NFWT is defined as the range of feedwater temperatures from NFWT to NFWT - 10F. NFWT and its allowable variation were assumed in the development of the Base Case Operation limits. The FHOOS limits and FFTR/Coastdown limits were developed for a maximum feedwater temperature reduction of 110.3F. The following must be considered when operating with RFWT:
Although the acronyms FWTR, FHOOS, RFWT and FFTR all involve reduced feedwater temperature, the use of FFTR is reserved for cycle energy extension using reduced feedwater temperature at and beyond a core average exposure of EOCLB using FFTR/Coastdown limits.
Prior to reaching the EOCLB exposure breakpoint, operation with FWTR >10F and reactor power 23% RTP requires use of the FHOOS limits.
Until a core average exposure of EOCLB is reached, implementation of the FFTR/Coastdown limits is not required even if coastdown begins early.
When operating with RFWT, the appropriate 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.
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. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 14 References In accordance with Brunswick Unit 1 Technical Specification 5.6.5.b, the analytical methods for determining Brunswick Unit 1 core operating limits have been specifically reviewed and approved by the NRC and are listed as References 1 through 21.
- 1. NEDE-24011-P-A, "GESTAR II - General Electric Standard Application for Reactor Fuel," and US Supplement, Revision 15, September 2005.
- 2. XN-NF-81-58(P)(A) and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Revision 2, March 1984.
- 3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Revision 1, September 1986.
- 4. EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Revision 0, February 1998.
- 5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs, Revision 1, May 1995.
- 6. XN-NF-80-19(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, March 1983.
- 7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Revision 1, June 1986.
- 8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Revision 0, October 1999.
- 9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX:
Thermal Limits Methodology Summary Description, Revision 2, January 1987.
- 10. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Revision 0, February 1987.
- 11. ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, June 2011.
- 12. ANF-913(P)(A) Volume 1 and Volume 1 Supplements 2, 3, 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Revision 1, August 1990.
- 13. ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors, Revision 3, September 2005.
- 14. EMF-2209(P)(A), SPCB Critical Power Correlation, Revision 3, September 2009.
- 15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Revision 0, August 2000.
- 16. EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model, Revision 0, May 2001.
- 17. EMF-2292(P)(A), ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients, Revision 0, September 2000.
- 18. EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2, Revision 0, August 2000.
- 19. NEDO-32465-A, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, August 1996.
- 20. BAW-10247PA, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, Revision 0, April 2008.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 15
- 21. ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014.
- 22. NEDC-31654P, Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, February 1989.
- 23. NEDO-31960-A, BWR Owners Group Long-Term Stability Solutions Licensing Methodology (Supplement 1), November 1995, NRC ADAMS Accession No. ML14093A211.
- 24. GENE-C51-00251-00-01, Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick 1 and 2, Revision 0, March 2001.
- 25. OG02-0119-260 Backup Stability Protection (BSP) for Inoperable Option III Solution, GE Nuclear Energy, July 17, 2002.
- 26. BAW-10255PA, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, Revision 2, May 2008.
- 27. BNP Design Calculation 1C51-0001, Power Range Neutron Monitoring System Setpoint Uncertainty and Scaling Calculation (1-C51-APRM-1 through 4 Loops and 1-C51 RBM-A and B Loops),
Revision 3, May 2004.
- 28. BNP Design Calculation 0B21-1015, BNP Power/Flow Maps, Revision 7, March 2008.
- 29. ANP-3636P, Brunswick Unit 1 Cycle 22 Reload Safety Analysis, Revision 0, December 2017.
- 30. BNP Design Calculation 1B21-2050, Preparation of the B1C22 Core Operating Limits Report, Revision 0, March 2018.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 16 Table 1 RBM System Setpoints1 a Setpoint Value Allowable Value Setpoint b < 27.7 < 29.0 Lower Power Setpoint (LPSP )
b < 62.7 < 64.0 Intermediate Power Setpoint (IPSP )
b < 82.7 < 84.0 High Power Setpoint (HPSP )
c,d < 117.1 < 117.6 Low Trip Setpoint (LTSP )
c,d < 112.3 < 112.8 Intermediate Trip Setpoint (ITSP )
c,d < 107.3 < 107.8 High Trip Setpoint (HTSP )
RBM Time Delay (td2) 0 seconds < 2.0 seconds a See Table 2 for RBM Operability Requirements.
b Setpoints in percent of Rated Thermal Power.
c Setpoints relative to a full scale reading of 125. For example, < 117.1 means
< 117.1/125.0 of full scale.
d Trip setpoints and allowable values are based on a HTSP Analytical Limit of 110.2 with RBM filter.
1 This table is referred to by Technical Specification 3.3.2.1 (Table 3.3.2.1-1) and 5.6.5.a.5.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 17 Table 2 RBM Operability Requirements2 IF the following conditions are met, THEN RBM Not Required Operable Thermal Power
(% rated) MCPR 1.89 TLO 29% and < 90%
Requirements valid for all fuel designs, all SCRAM insertion times and all core average exposure ranges.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 18 Table 3 PBDA Setpoints3 Amplitude Trip OLMCPR(SS) OLMCPR(2PT)
Setpoint (Sp) 1.05 1.16 1.19 1.06 1.18 1.20 1.07 1.19 1.22 1.08 1.21 1.24 1.09 1.23 1.26 1.10 1.25 1.28 1.11 1.27 1.30 1.12 1.29 1.32 1.13 1.31 1.34 1.14 1.33 1.36 1.15 1.35 1.39 Acceptance Criteria Off-rated OLMCPR @ Rated Power 45% Flow OLMCPR PDBA Setpoint Setpoint Value Amplitude Trip (Sp) 1.11 Confirmation Count (Np) 14 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 Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 19 Table 4 Exposure Basis4 for Brunswick Unit 1 Cycle 22 Transient Analysis Core Average Exposure Comments (MWd/MTU)
Breakpoint for exposure dependent MCPRp 34,284 limits (NEOC)
Design basis rod patterns to EOFP + 15 35,484 EFPD (EOCLB)
End of cycle with FFTR/Coastdown -
36,948 Maximum Core Exposure (MCE) 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 Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 20 Table 5 Power-Dependent MCPRp Limits5 NSS Insertion Times BOC to < NEOC EOOS Power ATRIUM 10XM
(%
Condition MCPRp rated) 100.0 1.35 90.0 1.37 Base 50.0 1.66 Case > 65%F 65%F Operation 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.38 90.0 1.40 50.0 1.66 TBVOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.35 90.0 1.37 50.0 1.66 FHOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20 5
Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 21 Table 6 Power-Dependent MCPRp Limits6 TSSS Insertion Times BOC to < NEOC EOOS Power ATRIUM 10XM
(%
Condition MCPRp rated) 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F Case Operation 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.41 90.0 1.44 50.0 1.66
> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.39 90.0 1.40 50.0 1.66
> 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20 6
Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 22 Table 7 Power-Dependent MCPRp Limits7 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM
(%
Condition MCPRp rated) 100.0 1.36 90.0 1.37 Base 50.0 1.66 Case > 65%F 65%F Operation 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.38 90.0 1.40 50.0 1.66 TBVOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.36 90.0 1.37 50.0 1.66 FHOOS > 65%F 65%F 50.0 1.91 1.78 26.0 2.34 2.22 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.38 90.0 1.40 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.91 1.78 26.0 2.34 2.22 26.0 3.03 2.96 23.0 3.22 3.20 7
Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 23 Table 8 Power-Dependent MCPRp Limits8 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM
(%
Condition MCPRp rated) 100.0 1.39 90.0 1.40 50.0 1.66 Base > 65%F 65%F Case Operation 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.38 2.34 23.0 2.45 2.43 100.0 1.41 90.0 1.44 50.0 1.66
> 65%F 65%F TBVOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.94 2.85 23.0 3.14 3.05 100.0 1.39 90.0 1.40 50.0 1.66
> 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 2.51 2.46 23.0 2.60 2.59 100.0 1.41 90.0 1.44 TBVOOS 50.0 1.66 and > 65%F 65%F FHOOS 50.0 1.93 1.80 26.0 2.36 2.25 26.0 3.03 2.96 23.0 3.22 3.20 8
Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 24 Table 9 Power-Dependent MCPRp Limits9 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 10XM
(%
Condition MCPRp rated)
Base Case 100.0 1.37 Operation 90.0 1.37 50.0 1.66 (FFTR/FHOOS > 65%F 65%F included) 50.0 1.91 1.78 26.0 2.34 2.22 (Bounds 26.0 2.51 2.46 operation with NFWT) 23.0 2.60 2.59 TBVOOS 100.0 1.38 90.0 1.40 (FFTR/FHOOS 50.0 1.66 included) > 65%F 65%F 50.0 1.91 1.78 (Bounds 26.0 2.34 2.22 operation with NFWT) 26.0 3.03 2.96 23.0 3.22 3.20 9
Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 25 Table 10 Power-Dependent MCPRp Limits10 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 10XM
(%
Condition MCPRp rated)
Base Case 100.0 1.39 Operation 90.0 1.40 50.0 1.66 (FFTR/FHOOS > 65%F 65%F included) 50.0 1.93 1.80 26.0 2.36 2.25 (Bounds 26.0 2.51 2.46 operation with NFWT) 23.0 2.60 2.59 TBVOOS 100.0 1.41 90.0 1.44 (FFTR/FHOOS 50.0 1.66 included) > 65%F 65%F 50.0 1.93 1.80 (Bounds 26.0 2.36 2.25 operation with NFWT) 26.0 3.03 2.96 23.0 3.22 3.20 10 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service. For single-loop operation, the TLO MCPRp limits shown above must be adjusted by adding 0.02. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 26 Table 11 Flow-Dependent MCPRf Limits11 Core Flow ATRIUM 10XM
(% of rated) MCPRf 0.0 1.55 31.0 1.55 60.0 1.47 80.0 1.30 100.0 1.30 107.0 1.30 11 Limits valid for all SCRAM insertion times, all core average exposure ranges, all EOOS scenarios, and both TLO
& SLO.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 27 Table 12 AREVA Fuel Steady-State LHGRSS Limits Peak ATRIUM 10XM Pellet Exposure LHGR (GWd/MTU) (kW/ft) 0.0 15.1 6.0 14.1 18.9 14.1 54.0 10.6 74.4 5.4
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 28 Table 13 AREVA Fuel Power-Dependent LHGRFACp Multipliers12 NSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM
(%
Condition LHGRFACp rated) 100.0 1.00 90.0 1.00 50.0 1.00 Base > 65%F 65%F Case Operation 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.51 0.53 23.0 0.49 0.50 100.0 1.00 90.0 1.00 50.0 1.00
> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.42 0.48 23.0 0.39 0.43 100.0 1.00 90.0 1.00 50.0 0.97
> 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 TBVOOS 50.0 0.96 and > 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40 12 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 29 Table 14 AREVA Fuel Power-Dependent LHGRFACp Multipliers13 TSSS Insertion Times BOC to < EOCLB EOOS Power ATRIUM 10XM
(%
Condition LHGRFACp rated) 100.0 1.00 90.0 1.00 50.0 1.00 Base > 65%F 65%F Case Operation 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.51 0.53 23.0 0.49 0.50 100.0 1.00 90.0 1.00 50.0 1.00
> 65%F 65%F TBVOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.42 0.48 23.0 0.39 0.43 100.0 1.00 90.0 1.00 50.0 0.97
> 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.46 0.48 23.0 0.44 0.46 100.0 1.00 90.0 1.00 TBVOOS 50.0 0.96 and > 65%F 65%F FHOOS 50.0 0.89 0.95 26.0 0.63 0.77 26.0 0.39 0.44 23.0 0.37 0.40 13 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 30 Table 15 AREVA Fuel Power-Dependent LHGRFACp Multipliers14 NSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 10XM
(%
Condition LHGRFACp rated)
Base Case 100.0 1.00 Operation 90.0 1.00 50.0 0.97 (FFTR/FHOOS > 65%F 65%F included) 50.0 0.89 0.95 26.0 0.63 0.77 (Bounds 26.0 0.46 0.48 operation with NFWT) 23.0 0.44 0.46 TBVOOS 100.0 1.00 90.0 1.00 (FFTR/FHOOS 50.0 0.96 included) > 65%F 65%F 50.0 0.89 0.95 (Bounds 26.0 0.63 0.77 operation with NFWT) 26.0 0.39 0.44 23.0 0.37 0.40 14 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 31 Table 16 AREVA Fuel Power-Dependent LHGRFACp Multipliers15 TSSS Insertion Times BOC to < MCE (FFTR/Coastdown)
EOOS Power ATRIUM 10XM
(%
Condition LHGRFACp rated)
Base Case 100.0 1.00 Operation 90.0 1.00 50.0 0.97 (FFTR/FHOOS > 65%F 65%F included) 50.0 0.89 0.95 26.0 0.63 0.77 (Bounds 26.0 0.46 0.48 operation with NFWT) 23.0 0.44 0.46 TBVOOS 100.0 1.00 90.0 1.00 (FFTR/FHOOS 50.0 0.96 included) > 65%F 65%F 50.0 0.89 0.95 (Bounds 26.0 0.63 0.77 operation with NFWT) 26.0 0.39 0.44 23.0 0.37 0.40 15 Limits support operation with any combination of any 1 inoperable SRV, 1 inoperable TBV, up to 40% of the TIP channels out-of-service, and up to 50% of the LPRMs out-of-service.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 32 Table 17 AREVA Fuel Flow-Dependent LHGRFACf Multipliers16 Core Flow ATRIUM 10XM
(% of rated) LHGRFACf 0.0 0.58 31.0 0.58 75.0 1.00 107.0 1.00 16 Multipliers valid for all SCRAM insertion times and all core average exposure ranges.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 33 Table 18 AREVA Fuel Steady-State MAPLHGRSS Limits17, 18 Average Planar ATRIUM 10XM Exposure MAPLHGR (GWd/MTU) (kW/ft) 0.0 13.1 15.0 13.1 67.0 7.7 17 AREVA Fuel MAPLHGR limits do not have a power, flow, or EOOS dependency.
18 ATRIUM 10XM MAPLHGR limits must be adjusted by a 0.80 multiplier when in SLO. SLO not permitted for FHOOS, TBVOOS or MSIVOOS.
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 34 Figure 1 Stability Option III Power/Flow Map OPRM Operable, Two Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 I
110.0 1-------,- -~J-:,-,-+,-P::-,-,-,-
APRM STP Rod Block 100 76.19 80.47 100.0 i-,I------ -- 99 98 75.04 73.89 80.47 80.47 i 71 I I
rrn TT 97 72.75 80.47 96 71.61 80.47 T1 rrrn 90.0 --,-,-,--n- -,-,-,-,-h-,- -,---,,. ,.,..-
I
, 0 00I - ,- ,- .- ,- ,- ,- ,- __ J _I_I_I_ O
_ - -n-,- J-~i.* --H---- i_:__
95 94 70.49 69.36 80.47 80.47 T 93 68.25 80.47 I 92 67.13 80.47 80.0 I rT
--1-1 JJ_
11 I 17 11
_1_1JJ__1_1_
17 JJ__1_1_1__
11 I
--n- +:JJ_i___
11 ITI I __1_
I , ITT T J__I_I_I_J_ 91 90 89 66.03 64.93 63.83 80.47 80.47 80.47 88 62.74 80.47 I ii I 87 61.66 80.51 70.0 - 1:i - _1_;~.. Ill _1_ H- --::: ,J,'- 86 60.58 80.60 I I 85 59.50 80.69 84 58.43 58 43 80.79 80 79
% Powerr 83 57.37 80.90 60.0 *,*+ 82 81 56.31 55.25 81.05 81.21 I 80 54.20 81.36
- 7~=:r::: ,~
79 53.16 81.51
+
1111 R
50.0 78 77 52.12 51.08 81.67 81.82 MELLL Line I e 76 50.05 81.98 T Rod Line SLO Entry C g 75 49.02 82.13 40.0 -H-H-H-H-H-1
_1_1_1_1 7 liii I I I I I I I I n 777-I I F i 74 73 48.00 46.98 82.29 82.44 o T 72 45.96 82.60 71 44.95 82.75
... ---- ----.---- -------- n
+
Scram Avoidance Region 70 43.94 82.91 30.0 I Il l 69 42.94 83.06 68 41.94 83.22
' 1111 N ' 67 40.95 83.37
-H-:+
66 39.96 83.52 20.0 -,- 65 38.97 83.68 64 37.99 83.83 Natural 1 11
' 63 37.01 83.99
+
62 36.04 84.14 10.0 Circulation -,-,-,-P.*.' -,-,-,-~ f-1 OPRM Enabled Region -,-,-,-:-.- H- 61 35.06 84.30 Line ,,., 60 34.10 84.45 I r, In 7 35% Minimum Pump Speed ' Minimum Power Line 59 58 33.13 32.17 84.61 84.70 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-1015, Revision 7
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 35 Figure 2 Stability Option III Power/Flow Map OPRM Inoperable, Two Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0
110.0
APRM STP Rod Block 100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 90.0 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 85 60.58 59.50 80.60 80.69 84 58.43 58 43 80.79 80 79
% Power 83 57.37 80.90 60.0 82 81 56.31 55.25 81.05 81.21 80 54.20 81.36 79 53.16 81.51 R
50.0 I e 78 77 52.12 51.08 81.67 81.82 MELLL Line 76 50.05 81.98 SLO Entry Rod Line C g 75 49.02 82.13 74 48.00 82.29 40.0 Region A - Manual SCRAM F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Region B - Immediate Exit n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 5% Buffer Region 66 39.96 83.52 20.0 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-1015, Revision 7
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 36 Figure 3 Stability Option III Power/Flow Map OPRM Operable, Single Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0
110.0
100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 APRM 97 72.75 80.47 96 71.61 80.47 STP Rod 95 70.49 80.47 90.0 Block 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 85 60.58 59.50 80.60 80.69 84 58 43 58.43 80 79 80.79
% Power 83 57.37 80.90 60.0 82 81 56.31 55.25 81.05 81.21 80 54.20 81.36 79 53.16 81.51 R
50.0 78 77 52.12 51.08 81.67 81.82 MELLL Line I e 76 50.05 81.98 C g 75 49.02 82.13 74 48.00 82.29 40.0 F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Scram Avoidance Region n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 66 39.96 83.52 20.0 45 Mlb/hr Max Core Flow 65 38.97 83.68 64 37.99 83.83 Natural Natural 63 37.01 83.99 62 36.04 84.14 Circulation 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 Line Line 60 34.10 84.45 59 33.13 84.61 35% Minimum 35% Minimum Pump Pump Speed Speed Minimum Power Minimum Line Power Line 58 32.17 84.70 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-1015, Revision 7
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 37 Figure 4 Stability Option III Power/Flow Map OPRM Inoperable, Single Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0
110.0
100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 APRM 97 72.75 80.47 96 71.61 80.47 STP Rod 95 70.49 80.47 90.0 Block 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 85 60.58 59.50 80.60 80.69 84 58 43 58.43 80 79 80.79
% Power 83 57.37 80.90 60.0 82 81 56.31 55.25 81.05 81.21 80 54.20 81.36 79 53.16 81.51 R
50.0 78 77 52.12 51.08 81.67 81.82 MELLL Line I e 76 50.05 81.98 C g 75 49.02 82.13 74 48.00 82.29 40.0 Region A - Manual SCRAM F i 73 46.98 82.44 o 72 45.96 82.60 71 44.95 82.75 Region B - Immediate Exit n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 5% Buffer Region 67 40.95 83.37 66 39.96 83.52 20.0 45 Mlb/hr Max Core Flow 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-1015, Revision 7
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 38 Figure 5 Stability Option III Power/Flow Map OPRM Operable, FWTR, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 I I I I
-~
110.0 _1_
I APRM STP Rod Block 100 76.19 80.47 100.0 + ..!-
-- -- -,- 99 98 97 75.04 73.89 72.75 80.47 80.47 80.47 I I I 11 1 96 71.61 80.47 I I I I 95 70.49 80.47 90.0 .I. [f_i_
94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 --,-
- I 90 89 64.93 63.83 80.47 80.47 88 62.74 80.47 87 61.66 80.51 70.0 -r ;- ---J.- -r 86 60.58 80.60 I I I 85 59.50 80.69 84 58.43 58 43 80.79 80 79
% Power 83 57.37 80.90 60.0 -,-,--,-,-h- -,-,-,-,--,-,- --,-,-,-,-,-- r,-h-,+ -------T-
- 7J-82 56.31 81.05 81 55.25 81.21 TT 11 IT I ill 11 1 I ITT I ,_.., T 80 54.20 81.36 79 53.16 81.51 50.0 MELLL Line R
I e + 78 77 76 52.12 51.08 50.05 81.67 81.82 81.98
_____~R+l=~~:t:t~~tt~=t+/-ttl:+/-+/-1 C g SLO Entry Rod Line 75 49.02 82.13
+
1
V- ---
74 48.00 82.29 40.0 (SLO prohibited F i 73 46.98 82.44
- /---- i during FWTR) o I 72 45.96 82.60 71 44.95 82.75
+
Scram Avoidance Region n 70 43.94 82.91 30.0 --------- 69 42.94 83.06 68 41.94 83.22 I I 67 40.95 83.37 66 39.96 83.52 20.0 J_ 65 38.97 83.68 64 37.99 83.83 Natural ' 63 62 37.01 36.04 83.99 84.14 10.0 Circulation -,-,-,-- OPRM Enabled Region -,- 61 60 35.06 34.10 84.30 84.45 Line 1 ,.
I I I 59 33.13 84.61
- I 35% Minimum Pump Speed I I Minimum Power Line 58 32.17 84.70 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-1015, Revision 7
Duke Energy, Nuclear Fuels Engineering, Nuclear Fuel Design Design Calc. No. 1B21-2050 Rev. 0 B1C22 Core Operating Limits Report, BNEI-0400-0017 Rev. 0 Page 39 Figure 6 Stability Option III Power/Flow Map OPRM Inoperable, FWTR, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0
110.0
APRM STP Rod Block 100 76.19 80.47 99 75.04 80.47 100.0 98 73.89 80.47 97 72.75 80.47 96 71.61 80.47 95 70.49 80.47 90.0 94 69.36 80.47 93 68.25 80.47 92 67.13 80.47 91 66.03 80.47 80.0 90 64.93 80.47 89 63.83 80.47 88 62.74 80.47 87 61.66 80.51 70.0 86 85 60.58 59.50 80.60 80.69 84 58.43 58 43 80.79 80 79
% Power 83 57.37 80.90 60.0 82 81 56.31 55.25 81.05 81.21 80 54.20 81.36 79 53.16 81.51 R
50.0 78 77 52.12 51.08 81.67 81.82 MELLL Line I e 76 50.05 81.98 SLO Entry Rod Line C g 75 49.02 82.13 74 48.00 82.29 40.0 Region A - Manual SCRAM (SLO prohibited F i 73 46.98 82.44 during FWTR) o 72 45.96 82.60 71 44.95 82.75 Region B - Immediate Exit n 70 43.94 82.91 30.0 69 42.94 83.06 68 41.94 83.22 67 40.95 83.37 5% Buffer Region 66 39.96 83.52 20.0 65 38.97 83.68 64 37.99 83.83 63 37.01 83.99 Natural 62 36.04 84.14 10.0 Circulation OPRM Enabled Region 61 35.06 84.30 60 34.10 84.45 Line 59 33.13 84.61 35% Minimum Pump Speed Minimum Power Line 58 32.17 84.70 0.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbs/hr Core Flow 0 10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
Reference:
0B21-1015, Revision 7