ML11319A024
| ML11319A024 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/10/2011 |
| From: | James Shea Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TVA-COLR-BF2C17, Rev 1, TVA-COLR-BF3C15, Rev 5 | |
| Download: ML11319A024 (76) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 November 10, 2011 10 CFR 50.4 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 2 and 3 Facility Operating License Nos. DPR-52 and DPR-68 NRC Docket Nos. 50-260 and 50-296
Subject:
Browns Ferry Nuclear Plant, Unit 2, Core Operating Limits Report for Cycle 17 Operation and Unit 3, Core Operating Limits Report for Cycle 15 Operation Revisions In accordance with the requirements of Technical Specification 5.6.5.d, the Tennessee Valley Authority is submitting the Unit 2 Cycle 17, Core Operating Limits Report (COLR)
Revision 1 and the Unit 3 Cycle 15, COLR, Revision 2. These revisions correct an editorial error that was in the previous revisions of the Unit 2 and Unit 3 COLRs, which were submitted to the NRC on March 2, 2011, and April 21, 2010 respectively.
There are no new commitments contained in this letter. If you have any questions please contact Tom Hess at (423) 751-3487.
Res c ully, W Sh e a
Enclosures:
- 1. Browns Ferry Unit 2 Cycle 17, Core Operating Limits Report, (105% OLTP), TVA-COLR-BF2C17, Revision 1 (Final)
- 2. Browns Ferry Unit 3 Cycle 15, Core Operating Limits Report, (105% OLTP), TVA-COLR-BF3C15, Revision 2 (Final) cc: See Page 2
U.S. Nuclear Regulatory Commission Page 2 November 10, 2011 cc: (w/ Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant
Enclosure I Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 2 Cycle 17, Core Operating Limits Report, (105% OLTP),
TVA-COLR-BF2CI7, Revision I (Final)
(See Attached)
EDMS L32 110906 800 QA Document Pages Affected: All BFE-3086, Revision 1 ETNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga, TN 37402 Browns Ferry Unit 2 Cycle 17 Core Operating Limits Report, (105% OLTP)
TVA-COLR-BF2C17 Revision 1 (Final)
(Revision Log, Page v)
September 2011 Prepared:
-I. W. Eichenb-erg, Sr. Sp/cialist Date:
_.,. W/ 4 2-.'1 Date:
6/
-4, ý__ 1 ýýý ý --, ---
Verified:
B.-
litchell, Engineer
-Approved:
6 r
Date:
G. C. Storey, Man ger, BWR Fuel Engineering Reviewed:
Approved:
D*
W. R. Hayes, Mana46b eactor Engineering I-'
ite:
ite:
LJC Chairman, PORC
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 N PG 1101 Market Street, Chattanooga TN 37402 Table of Contents Total Number of Pages = 39 (including review cover sheet)
L ist o f T a b le s.................................................................................................................................................
III L ist o f F ig u re s..........................................
iv Revision Log...................................................................................................................................
v N o m e n c la tu re..................................................................................................................................................
v i References...................................................................................................................................................
viii 1
Introduction..........................................................................................................................
1 1.1 Purpose.........................................................................................................................
1 1.2 S c o p e............................................................................................................................
1 1.3 Fuel Loading..................................................................................................................
1 1.4 Acceptability..................................................................................................................
1 2
APLHGR Lim its....................................................................................................................
3 2.1 Rated Power and Flow Lim it: APLHG RRATED..........................................................
3 2.2 Off-Rated Power Dependent Lim it: APLHG Rp.........................................................
3 2.2.1 Startup w ithout Feedwater Heaters.................................................................
3 2.3 Off-Rated Flow Dependent Lim it: APLHG RF...........................................................
3 2.4 Single Loop O peration Lim it: APLHG RSLO...............................................................
3 2.5 Equipm ent O ut-Of-Service Corrections.....................................................................
5 3
LHG R Lim its.........................................................................................................................
6 3.1 Rated Power and Flow Lim it: LHG RRATED.......................................
............................. 6 3.2 Off-Rated Power Dependent Lim it: LHG Rp..............................................................
6 3.2.1 Startup w ithout Feedwater Heaters...................................................................
6 3.3 Off-Rated Flow Dependent Lim it: LHG RF.................................................................
7 3.4 Equipm ent O ut-Of-Service Corrections.....................................................................
7 4
O LM CPR Lim its.................................................................................................................
13 4.1 Flow Dependent M CPR Lim it: M CPRF...................................................................
13 4.2 Power Dependent M CPR Lim it: M CPRp................................................................
13 4.2.1 Startup without Feedwater Heaters................................................................
13 4.2.2 Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS).............................
14 4.2.3 Exposure Dependent Lim its............................................................................
14 4.2.4 Equipm ent O ut-Of-Service (EO OS) O ptions..................................................
15 4.2.5 Single-Loop-Operation (SLO) Limits........
- ..................... 15 4.2.6 Below Pbypass Lim its.....................................................................................
15 5
APRM Flow Biased Rod Block Trip Settings
.......... 24 6
Rod Block M onitor (RBM ) Trip Setpoints and O perability..............................................
25 7
Shutdown Margin Limit................................................................ 27 Appendix A:
Therm al-Hydraulic Stability......................................
28 BrtMns Fny Uit 2cyde17IP geII Browns Ferry Unit 2 cycle 17 Core Oper'abrng Limits Report, (105% OLTP)
Page ii TVA-C0LR-I3F2C17, Revision 1 (Final)
EDMS: L32 110906 800 EIMNPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 List of Tables N uclear Fuel T ypes..........................................................................................................
.... 2 Startup Feedwater Temperature Basis.....................................................................................
6 Nominal Scram Time Basis.............................................
14 MCPRp Limits for Optimum Scram Time Basis.....................................................................
17 MCPRp Limits for Nominal Scram Time Basis.......................................................................
18 MCPRp Limits for Technical Specification Scram Time Basis.......................
20 Startup Operation MCPRp Limits for Table 3.1 Temperature Range 1:
Technical Specification Scram Time Basis...........................................................................
22 Startup Operation MCPRp Limits for Table 3.1 Temperature Range 2:
Technical Specification Scram Time Basis............................................................................
23 Analytical RBM Trip Setpoints..............................................................................................
25 R BM S etpoint A pplicability......................................................................................................
25 Control Rod Withdrawal Error Results...................................................................................
26 O P R M S etpo ints..........................................................................................................................
30 B~tMns FriyUnit2 C~e 1 Pag iI Browns Ferry Unit 2 Cy (Re 17 Core Operating Limit Report,( 105% OLTP)
Page iii TVA-C0LR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date:
September 12, 2011 1101 Market Street, Chattanooga TN 37402 List of Figures APLHG RRATED for ATRIUM -10 Fuel........................................................................................
4 LHGRRATED for ATRIUM-10 Fuel................................................
8 Base Operation LHGRFACp for ATRIUM-1 0 Fuel..................................................................
9 LHGRFACF for ATRIUM-1 0 Fuel............................................
10 Startup Operation LHGRFACp for ATRIUM-10 Fuel: Table 3.1 Temperature Range 1......
11 Startup Operation LHGRFACp for ATRIUM-10 Fuel: Table 3.1 Temperature Range 2......
12 M C PR F for A TR IU M -10 Fuel.....................................................................................................
16 Browns Feny Unit 2 Cyde 17 Core Operating Umits Report, (105% OLTP)
Page iv TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 IM ~NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Revision Log I
I I Number I Page 1-R1 0-RO 9,111,'&
12 All Description Editorial fix. Correct X-axis label to read: Core Power (% Rated)
New document
-j Browns FerryUnft2 Cyde 17 Core Operaling LimitsReport (105% OLTP)
Page v TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 EIMNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Date: September 12, 2011 Nomenclature APLHGR APRM AREVA NP Average Planar LHGR Average Power Range Monitor Vendor (Framatome, Siemens)
Beginning of Cycle Backup Stability Protection Boiling Water Reactor BOC BSP BWR CAVEX CD CMSS COLR CPR CRWE CSDM DIVOM EOC EOOS FFTR FFWTR FHOOS ft GWd HTSP ICA ICF IS kW Core Average Exposure Coast Down Core Monitoring System Software Core Operating Limits Report Critical Power Ratio Control Rod Withdrawal Error Cold SDM Delta CPR over Initial CPR vs. Oscillation Magnitude End of Cycle Equipment OOS Final Feedwater Temperature Reduction Final Feedwater Temperature Reduction Feedwater Heaters OOS Foot: english unit of measure for length Giga Watt Day High TSP Interim Corrective Action Increased Core Flow (beyond rated)
In-Service kilo watt: SI unit of measure for power.
License Condition of Operation Loss of Feedwater Heating, LHGR Multiplier (Power or Flow dependent)
Low Power Range Monitor Generator Load Reject, No Bypass MAPLHGR multiplier (Power or Flow dependent)
Minimum CPR Moisture Separator Reheater Valve LCO LFWH LHGRFAC LPRM LRNB MAPFAC MCPR MSRV Browns Ferry Unit 2 Cyde 17 Core Operating i~mits.Report, (105% OLTP)
Page vi TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date:
September 12, 2011 N PG
~~~110 1 Market Street, Chattanooga TN 37402 Dt:Spebr1,21 MSRVOOS MSRV OOS MTU Metric Ton Uranium MWd/MTU Mega Watt Day per Metric Ton Uranium NEOC Near EOC NRC United States Nuclear Regulatory Commission NSS Nominal Scram Speed NTSP Nominal TSP OLMCPR MCPR Operating Limit OOS Out-Of-Service OPRM Oscillation Power Range Monitor OSS Optimum Scram Speed PBDA Period Based Detection Algorithm Pbypass Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU Power Load Unbalance PLUOOS PLU OOS PRNM Power Range Neutron Monitor RBM Rod Block Monitor RPS Reactor Protection System RPT Recirculation Pump Trip RPTOOS RPT OOS SDM Shutdown Margin SLMCPR MCPR Safety Limit SLO Single Loop Operation TBV Turbine Bypass Valve TBVIS TBV IS TBVOOS Turbine Bypass Valves OOS TIP Transversing In-core Probe TIPOOS TIP OOS TLO Two Loop Operation TSP Trip Setpoint TSSS Technical Specification Scram Speed TVA Tennessee Valley Authority Browns Ferry Unit2 Cyde 17 Page vii Core Operating Umits Report (105% OLTP)
"VA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date:
September 12, 2011 NPG 1101 Market Street, Chattanooga TN 37402 References
- 1.
ANP-2964, Revision 0, Browns Ferry Unit 2 Cycle 17 Reload Safety Analysis, AREVA NP, Inc., December, 2010.
- 2.
ANP-2537(P) Revision 0, Mechanical Design Report for Browns Ferry Unit 2 Reload BFE2-15 ATRIUM-10 Fuel Assemblies, AREVA NP, Inc., May 2006.
- 3.
ANP-2755(P) Revision 0, Mechanical Design Report for Browns Ferry Unit 2 Reload BFE2-16 ATRIUMTM-10 Fuel Assemblies, AREVA NP, Inc., November 2008.
- 4.
ANP-2939(P), Rev. 0, Mechanical Design Report for Browns Ferry Unit 2 Reload BFE2-17 ATRIUM-10 Fuel Assemblies, AREVA NP, Inc., July 2010.
- 5.
ANP-2913(P) Revision 1, Browns Ferry Unit 2 Cycle 17 Plant Parameters Document, AREVA NP, Inc., November 2010.
- 6.
BFE-3083, Revision 0, Verification of Browns Ferry Unit 2 Reload 16 Cycle 17 InCore Shuffle, Calculation File, Tennessee Valley Authority, February 2011.
Methodology References
- 7.
XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
- 8.
XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
- 9.
EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
- 10.
ANF-89-98(P)(A) Revision land Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 11.
XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
- 12.
XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
- 13.
EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
- 14.
XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
Brwns Ferry Unit 2 Cyce 17 Core Operatng Umits Report, (105% OLTP)
Page viii TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 NPG 1101 Market Street, Chattanooga TN 37402
- 15.
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, Exxon Nuclear Company, February 1987.
- 16.
ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.
- 17.
ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
- 18.
ANF-1358(P)(A) Revision 1, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 1992.
- 19.
EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, Siemens Power Corporation, September 2009.
- 20.
EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001.
- 21.
EMF-2292(P)(A) Revision 0, ATRIUM TM-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 22.
EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
- 23.
BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, Framatome ANP, Inc., May, 2008.
PRNM Setpoint References.
- 24.
Filtered Setpoints - EDE-28-0990 Rev. 3 Supplement E, "PRNM (APRM, RBM, and RFM) Setpoint Calculations [ARTS/MELLL (NUMAC) - Power-Uprate Condition] for Tennessee Valley Authority Browns Ferry Nuclear Plant", October 1997.
- 25.
Unfiltered Setpoints - EDE-28-0990 Rev. 2 Supplement E, "PRNM (APRM, RBM, and RFM) Setpoint Calculations [ARTS/MELLL (NUMAC) - Power-Uprate Condition]
for Tennessee Valley Authority Browns Ferry Nuclear Plant", October 1997.
- 26.
GE Letter LB#: 262-97-133, Browns Ferry Nuclear Plant Rod Block Monitor Setpoint Clarification - GE Proprietary Information, September 12, 1997.
- 27.
NEDC-32433P, Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2, and 3, GE'Nuclear Energy, April 1995.
Browns Ferry Unft 2 cyde 17 Core Operating Urnits Report (105% OLTP)
Page ix TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 NPG 1101 Market Street, Chattanooga TN 37402 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1 Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2 Scope This document will discuss the following areas:
Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
> Minimum Critical Power Ratio Operating Limit (OLMCPR)
-(Technical Specifications 3.2.2, 3.3.4.1, and 3.7.5)
> Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
> Shutdown Margin (SDM) Limit (Technical Specification 3.1.1) 1.3 Fuel Loading The core will contain all AREVA NP, Inc., ATRIUM-10 fuel. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented in Reference 6.
1.4 Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 7 through 23.
Browns Fery Unit 2 Cyde 17 Core Operaltng Limi ReporL, (105% OLTP)
Page 1 TVA-COLR-BF2C17, Revision I (Final)
EDMS: L32 110906 800 Date: September 12, 2011 Eu NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 1.1 Nuclear Fuel Types Nuclear' Original Number of Fuel Type Fuel Names Fuel Description Cycle Assemblies (NFT)
(Range)
ATRIUM-10 A10-4227B-15GV80-FBB 15 112 2
FBB001-FBB206 ATRIUM-10 A10.-4239B-15GV80-FBB 15 81 3
FBB207-FBB317 ATRIUM-10 A10-3552B-10GV80-FBB 15 16 4
FBB319-FBB374 ATRIUM-lo A10-4218B-13GV80-FCC 16 16 6
FCC291 -FCC306 ATRIUM-10 A10-3757B-1OGV80-FCC 16 24 7
FCC311 -FCC334 ATRIUM-10 A10-4019B-14GV80-FBC 16 167 8
FBC401-FBC568 ATRIUM-10 A10-3841B-14GV80-FBC 16 76 9
FBC569-FBC644 ATRIUM*10 A10-3799B-14GV80-FBD 17 136 10 FBDO01-FBD136 ATRIUM-10 A10-4004B-15GV80-FBD 17 136 11 FBD137-FBD272 The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends up6n uncertainties during the outage such as discovering a failed fuel bundle, or otherbundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 2 Cycle 17 CoeOperating Limits Report, (105%Y OLTP)
Page 2 TVA-C0LR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date:
September 12, 2011 NPG 1101 Market Street, Chattanooga TN 37402 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN (APLHGRp, APLHGRF, APLHGRsLo) where:
APLHGRp off-rated power APLHGR limit
[APLHGRRATED
- MAPFACp]
APLHGRF off-rated flow APLHGR limit
[APLHGRRATED
- MAPFACF]
[APLHGRRATED
- SLO Multiplier]
2.1 Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR, for all fuel types, is identified in Reference 1 and shown in Figure 2.1.
2.2 Off-Rated Power Dependent Limit: APLHGRp Reference 1, for ATRIUM-1 0 fuel, does not specify a power dependent APLHGR. Therefore, MAPFACP is set to a value of 1.0.
2.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reach a significant operating power level. No Additional power dependent limitation is required.
2.3 Off-Rated Flow Dependent Limit: APLHGRF Reference 1, for ATRIUM-10 fuel, does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4 Single Loop Operation Limit: APLHGRSLO The single loop operation multiplier for ATRIUM-1 0 fuel is 0.85, per Reference 1.
Browns Feny Unit 2Cydel17 CoreOperating Umits Report. (105% OLTP).
Page 3 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 EIM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 15 12
-J O..6 3
0 0
20 40 60 PlanarAverage Exposure (GWd/MTU) 80 Planar Avg.
Exposure APLHGR Limit (GWd/MTU)
(kW/ft) 0.0 12.5 15.0 12.5 67.0 7.3 Figure 2.1 APLHGRRATED for ATRIUM-10 Fuel Browns Feny Unit 2 Cyde 17 Core Operating Umits Report (105% OLTP)
Page 4 rVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 IM ~NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 2.5 Equipment Out-Of-Service Corrections The limit shown in Figure 2.1 is applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service RPTOOS TBVOOS PLUOOS FHOOS (or FFWTR)
All equipment In-Service (includes 1 SRVOOS)
EOC-Recirculation Pump Trip Out-Of-Service Turbine Bypass Valve(s) Out-Of-Service Power Load Unbalance Out-Of-Service Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction Single Recirculation Loop Operation (SLO) requires the application of the SLO multipliers to the rated APLHGR limits as described previously.
Browns Ferry Unit 2 C~yde 17 CoreOperatrng 1mits Repo~rt (105% OLTP)
Page 5 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 I NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 3
LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN ( LHGRp, LHGRF) where:
LHGRp LHGRF off-rated power LHGR limit off-rated flow LHGR limit
[LHGRRATED
- LHGRFACp]
[LHGRRATED
- LHGRFACF]
3.1 Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR, for all fuel types, is identified in Reference 1 and shown in Figure 3.1. The LHGR limit is consistent with References 2, 3, and 4.
3.2 Off-Rated Power Dependent Limit: LHGRp The ATRIUM-10 fuel, LHGR limits are adjusted for off-rated power conditions using the LHGRFACp multiplier provided in Reference 1. The multiplier is split into two sub cases:
turbine bypass valves in and out-of-service. The multipliers are shown in Figure 3.2.
3.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reach a significant operating power level. Additional limits are shown in Figure 3.4 and Figure 3.5, based on temperature conditions identified in Table 3.1.
Table 3.1 Startup Feedwater Temperature Basis Temperature Power Range 1 Range 2
(....t.ed)
(*F)
(oF) 25 160.0 155.0 30 165.0 160.0 40 175.0 170.0 50 185.0 180.0.
Browns Feny unt2 C~e 17 Page 6 Browns Ferry Unit 2 Cy (Fe 17 Core Operabdng Limits Report, (105% OLTP)
Page 6 TVA-C0LR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 EIM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 3.3 Off-Rated Flow Dependent Limit: LHGRF The ATRIUM-10 fuel, LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. The multiplier is shown in Figure 3.3.
3.4 Equipment Out-Of-Service Corrections The limit shown in Figure 3.1 is applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.*
In-Service RPTOOS TBVOOS PLUOOS FHOOS (or FFWTR)
SLO All equipment In-Service EOC-Recirculation Pump Trip Out-Of-Service Turbine Bypass Valve(s) Out-Of-Service Power Load Unbalance Out-Of-Service Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction Single Loop Operation, One Recirculation Pump Out--Of-Service Off-rated power corrections shown in Figure 3.2 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFWTR, or SLO.
Off-rated power corrections shown in Figure 3.4 and Figure 3.5 are also dependent on operation of the Turbine Bypass Valve system. In this case, limits support FHOOS operation during startup. These limits have no dependency on RPTOOS, PLUOOS, or SLO.
All equipment service conditions assume 1 SRVOOS.
Bn~ns Feny unit2 Cyde 17 Page 7 Browns Ferry UnRt 2 cycle 17 Core operating Umifts Report (105% OLTP)
Page 7 TVA-C0LR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800
[U.NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 15 12 0z
-J 9
6 3
0 0
20 40 60 Pellet Exposure (GWd/MTU) 80 Pellet LHGR Exposure Limit (GWd/MTU)
(kW/ft) 0.0 13.4 18.9 13.4 74.4 7.1 Figure 3.1 LHGRRATED for ATRIUM-10 Fuel Brawns FerTyUnit2Cyde 17 Page 8 Browns Ferry Unit 2 Cycle 17 Core Operating Limits Report, (105% OLTP)
Page 8 TVA-COLR-BF2C17, ReOsion 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 UNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 0.
x.
1.10 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 20 30 40 50 60 70 80 90 100 110 Core Power (% Rated)
Turbine Bypass In-Service Turbine Bypass Out-of-Service Core Power
(%Rated) 100.0 30.0 Core Flow 30.0 25.0 Core Flow 30.0 25.0 LHGRFACp TI 1.00 0.60 50% Rated 0.48 0.45 50% Rated 0.49 0.46 Core Power
(%Rated) 100.0 30.0 Core Flow 30.0 25.0 Core Flow 30.0 25.0 LHGRFACp 0.92 0.60
> 50% Rated S 0.42 0.38
< 50% Rated 0.49 0.43 Figure 3.2 Base Operation LHGRFACp for ATRIUM-10 Fuel (Independent of other EOOS conditions)
Browns Ferry Unit 2 Cyde 17 Core Operating ULmits Report, (105% OLTP)
Page 9 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 UNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 1.10 1.05 1.00 L-0,-I 0.95 0.90 0.85 0.80 30 40 50 60 70 80 90 100 Core Flow (% Rated) 110.
Core Flow LHGRFACF
(% Rated) 30.0 0.9 52.0 1
107.0 1
Figure 3.3 LHGRFACF for ATRIUM-10 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Fery Unit 2 Cyde 17 Core Operating Limits Report, (105% OLTP)
Page 10 TVA-COLR-BF2C17, Revsion 1 (Final)
EDMS: L32 110906 800 IM NPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 IL C,
1.10 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 20 30 40 50 60 70 80 90 100 110 Core Power (% Rated)
Turbine Bypass In-Service Core i
Power
(% Rated) 100.0 50.0 50.0
.30.0 Core Flow 30.0 25.0 Core Fow 30.0 25.0 1.00 0.71 0.68 0.57
> 50% Rated 0.42 0.39
-< 50% Rated 0.43 0.40
- LHGRFAC, Turbine Bypass Out-of-Service Core Power LHGRFACp
(%Rated) 4 I_
100.0 0.92 50.0 0.69 50.0 0.68 30.0 0.55 Core Flow > 50% Rated 30.0 F 0.40 25.0 I
0.36 Core Row <5 50% Rated 30.0 0.43 25.0
[
0.39 Figure 3.4 Startup Operation LHGRFACp for ATRIUM-10 Fuel:
Table 3.1 Temperature Range 1 (no Feedwater heating during startup) a Browns Ferry Unit 2 Cyde 17 Core Operating Umits Report (105% OLTP)
Page 11 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 E~IUNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 LL 1.10 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 20 30 40 50 60 70 80 90 100 110 Core Power (% Rated)
Turbine Bypass In-Service Core Power LHGRFACp
(% Rated) 100.0 1.00 50.0 0.71 50.0 0.68 30.0 0.55 Core Flow > 50% Rated 30.0
[
0.42 25.0 0.39 Core Row < 50% Rated 30.0 0.43 25.0 I
0.39' Turbine Bypass Out-of-Service Core Power LHGRFACp
(%Rated) 100.0 0.92 50.0 0.69 50.0 0.68 30.0 0.55 Core Fow > 50% Rated 30.0 1
0.39 25.0 J
0.35 Core Row < 50% Rated 30.0 0.43 25.0 0.39 Figure 3.5 Startup Operation LHGRFACp for ATRIUM-10 Fuel:
Table 3.1 Temperature Range 2 (no Feedwater heating during startup)
Browns Ferry Unit2 Cycde 17 Core Operating Limts Report (105% OLTP)
Page 12 lVA-COLR-8F2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 IM NPG 1101 Market Street, Chattanooga TN 37402 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF, MCPRP) where:
MCPRF core flow-dependent MCPR limit MCPRp power-dependent MCPR limit 4.1 Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2 Power Dependent MCPR Limit: MCPRp MCPRp limits are dependent upon:
" Core Power Level (% of Rated)
" Technical Specification Scram Speed (TSSS), Nominal Scram Speed (NSS), or Optimum Scram Speed (OSS)
" Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
Equipment Out-Of-Service Options
" Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRp limits are provided in the following tables, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range). The CMSS determines MCPRp limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Startup without Feedwater Heaters There is a range of operation during startup when the feedwater heaters are not placed into service until after the unit has reach a significant operating power level. Additional power dependent limits are shown in Table 4.5 and Table 4.6, based on temperature conditions identified in Table 3.1.
Browns FerryUn 2 Cyde 17 Core Operabng Umrt Report, (105% OLTP)
Page 13 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 EU NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 4.2.2 Scram Speed Dependent Limits (TSSS vs. NSS vs. OSS)
MCPRp limits are provided for three different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRp limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Both Nominal Scram Speeds (NSS) and/or Optimum Scram Speeds (OSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.*t Table 4.1 Nominal Scram Time Basis Notch Nominal Optimum Position Scram Timing Scram Timing (index)
(seconds)
(seconds)
I 46 0.420 0.380 36 0.980 0.875 26 1.600 1.465 6
2.900 2.900 In demonstrating compliance with the NSS and/or OSS scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRp limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS and/or OSS based limits are applicable.
4.2.3 Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRp limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRp limits are provided for the following exposure ranges:
BOC to NEOC BOC to EOC BOC to End of Coast NEOC corresponds to EOC corresponds to End of Coast 29,179.5 MWd / MTU 31,822.1 MWd / MTU 32,730.9 MWd / MTU NEOC refers to a Near EOC exposure point.
Reference 1 analysis results are based on information identified in Reference 5.
t Assumption basis is consistent with method used to perform actual timing measurements, (i.e., including pickup/dropout effects).
Browns Ferry Unit 2 Cycle 17 Core Operating Urmits Report (105% OLTP)
Page 14 "VA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 EIU NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 The EOC exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
4.2.4 Equipment Out-Of-Service (EOOS) Options EOOS options covered by MCPRp limits are given by the following:
In-Service RPTOOS TBVOOS RPTOOS+TBVOOS PLUOOS PLUOOS+RPTOOS PLUOOS+TBVOOS PLUOOS+TBVOOS+RPTOOS FHOOS (or FFWTR)
All equipment In-Service EOC-Recirculation Pump Trip Out-Of-Service Turbine Bypass Valve(s) Out-Of-Service Combined RPTOOS and TBVOOS Power Load Unbalance Out-Of-Service Combined PLUOOS and RPTOOS Combined PLUOOS and TBVOOS Combined PLUOOS, RPTOOS, and TBVOOS Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
For exposure ranges up to NEOC and EOC, additional combinations of MCPRp limits are also provided including FHOOS. The coast down exposure range assumes application of FFWTR.
FHOOS based MCPRP limits for the coast down exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.5 Sin-gle-Loop-Operation (SLO) Limits MCPRp limits are increased by 0.02 to support SLO, per Reference 1.
4.2.6 Below Pbypass Limits Below Pbypass (30% rated power), MCPRp limits depend upon core flow. One set of MCPRp limits applies for core flow above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
All equipment service conditions assume 1 SRVOOS.
Browns Feny Unit 2 cycle 17 Core Operating Umits Report (105% OLTP)
Page 15 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 IM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402
- 1 2.00 1.80 1.60 a-1.40 1.20 1.00 30 40 50 60 70 80 90 100 110 Core Flow (% Rated)
Core Flow MCPRF
(/(% Rated) 30.0 1.61 78.0 1.28 107.0 1.28 Figure 4.1 MCPRF for ATRIUM-10 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Browns Ferry Unit 2 Cyce 17 Core Operadng Limits Report, (105% OLTP)
Page 16 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011
[M NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 4.2 MCPRp Limits for Optimum Scram Time Basis*
BOC BOC BOC Power to to to End of Operating Condition
(% of rated)
NEOC EOC Coast 100 1.38 1.41 1.44 75 1.51 1.51 1.56 65 1.60 1.61 1.68 50 1.78 1.78 50 1.85 1.85 1.90 Base Case 40 2.00 2.00 2.13 30 2.29 2.29 2.44 30 at > 50%F 2.79 2.79 2.90 25 at > 50%F 3.08 3.08 3.22 30 at < 50%F 2.72 2.72 2.82--
25 at < 50%F 2.97 2.97 3.10
- All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR/FHOOS is supported for the BOC to End of Coast limits.
Browns Ferry Unit 2 Cycle 17 Core Operating Umits Report, (105% OLTP)
Page 17 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 U NPG Nuclear Fuel Engineering - BWRFE Date:
September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 4.3 MCPRp Limits for Nominal Scram Time Basis*
BOC BOC BOC Operating ower to to to End of Condition
(% of rated)
NEOC EOC Coast 100 1.41 1.42 1.44 75 1.52 1.53 1.56 65 1.62 1.62 1.68 50 1.82 1.82 50 1.86 1.86 1.90 Base Case 40 2.03 2.03 2.13 30 2.32 2.32 2.44 30 at > 50%F 2.79 2.79 2.90 25 at > 50%F 3.08 3.08 3.22 30 at < 50%F 2.72 2.72 2.82 25 at < 50%F 2.97 2.97 3.10 100 1.44 1.46 1.47 75 1.57 1.57 1.60 65 1.66 1.67 1.70 50 1.83 1.83 50 1.86 1.86 1.90 TBVOOS 40 2.04 2.04 2.13 30 2.32 2.32 2.44 30 at > 50%F 3.26 3.26 3.40 25 at > 50%F 3.70 3.70 3.85 30 at < 50%F 2.85 2.85 3.00 25 at S 50%F 3.29 3.29 3.47 100 1.43 1.44 75 1.55 1.56 65 1.68 1.68 50 50 1.90 1.90 FHOOS 40 2.13 2.13 30 2.44 2.44 30 at > 50%F 2.90 2.90 25 at > 50%F 3.22 3.22 30at S 50%F 2.82 2.82 25 at5 50%F 3.10 3.10 100 1.41 1.42 1.44 75 1.52 1.53 1.56 65 1.76 1.77 1.77 50 50 1.86 1.86 1.90 PLUOOS 40 2.03 2.03 2.13 30 2.32 2.32 2.44 30 at > 50%F 2.79 2.79 2.90 25 at > 50%F 3.08 3.08 3.22 30 at 5 50%F 2.72 2.72 2.82 1
25 at < 50%F 2.97 2.97 3.10 All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 2 Cyde 17 Core Operating Limits Report, (105% OLTP)
Page 18 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 V !~NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 4.3 MCPRp Limits for Nominal Scram Time Basis (continued)
BOC BOC BOC Pbwer to to to End of Operating Condition
(% of rated)
NEOC EOC Coast 100 1.46 1.47 75 1.58 1.59 65 1.70 1.70 50 rBVOOS 50 1.90 1.90 FHOOS 40 2.13 2.13 30 2.44 2.44 30 at > 50%F 3.40 3.40 25 at > 50%F 3.85 3.85 30 at: -50%F 3.00 3.00 25 at < 50%F 3.47 3.47 100 1.44 1.46 1.47 75 1.57 1.57 1.60 65 1.76 1.77 1.77 50 50 1.86 1.86 1.90 TBVOOS 40 2.04 2.04 2.13 P.UOOS 30 2.32 2.32 2.44 30 at > 50%F 3.26 3.26 3.40 25 at > 50%F 3.70 3.70 3.85 30 at - 50%F 2.85 2.85 3.00 25 at - 50%F 3.29 3.29.
3.47 100 1.43 1.44 75 1.55 1.56 65 1.76 1.77 50 FHOOS 50 1.90 1.90 40 2.13 2.13 30 2.44 2.44 30 at > 50%F 2.90 2.90 25 at > 50%F 3.22 3.22 30 at5 <50%F 2.82 2.82 25 at5 <50%F 3.10 3.10 100 1.46 1.47 75 1.58 1.59 65 1.76 1.77 50 TBVOOS 50 1.90 1.90 FHOOS 40 2.13 2.13 PLUOOS 30 2.44 2.44 30 at > 50%F 3.40 3.40 25 at > 50%F 3.85 3.85 30 at < 50%F 3.00 3.00 1
25 at < 50%F 3.47 3.47
" All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without
Page 19 TVA-COLR-BF2C17, Revision I (Final)
EDMS: L32 110906 800 IM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 4.4 MCPRp Limits for Technical Specification Scram Time Basis BOC BOC BOC Power to to to End of Operating Condition
(% of rated)
NEOC EOC Coast 100 1.43 1.43 1.45 75 1.54 1.54 1.58 65 1.65 1.65 1.72 50 1.85 1.85 50 1.87 1.87 1.93 Base Case 40 2.06 2.06 2.16 30 2.36 2.36 2.47 30 at > 50%F 2.79 2.79 2.90 25 at > 50%F 3.08 3.08 3.22 30 at < 50%F 2.72 2.72 2.82 25 at5 50%F 2.97 2.97 3.10 100 1.46 1.47 1.48 75 1.58 1.58 1.61 65 1.69 1.69 1.73 50 1.86 1.86 50 1.87 1.87 1.94 TBVOOS 40 2.07 2.07 2.16 30 2.36 2.36 2.47 30 at > 50%F 3.26 3.26 3.40 25 at > 50%F 3.70 3.70 3.85 30 at < 50%F 2.85 2.85 3.00 25 at < 50%F 3.29 3.29 3.47 100 1.45 1.45 75 1.58 1.58 65 1.72 1.72 50 50 1.93 1.93 FHOOS 40 2.16 2.16 30 2.47 2.47 30 at > 50%F 2.90 2.90 25 at > 50%F 3.22 3.22 30 at < 50%F 2.82 2.82 25 atS< 50%F 3.10 3.10 100 1.43 1.43 1.45 75 1.54 1.54 1.58 65 1.77 1.78 1.79 50 50 1.87 1.87 1.93 PLUOOS 40 2.06 2.06 2.16 30 2.36 2.36 2.47 30 at > 50%F 2.79 2.79 2.90 25 at > 50%F 3.08 3.08 3.22 30 at S 50%F 2.72 2.72 2.82 25 at < 50%F 2.97 2.97 3.10 All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequenty, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 2 Cycle 17 Core Operating Limits Report (105% OLTP)
Page 20 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 IM NPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 4.4 MCPRp Limits for Technical Specification Scram Time Basis (continued)*
BOC BOC BOC Power to to to End of Operating Condition
(% of rated)
NEOC EOC Coast 100 1.48 1.48 75 1.61 1.61 65 1.73 1.73 50 mVOOS 50 1.94 1.94 40 2.16 2.16 30 2.47 2.47 30 at > 50%F 3.40 3.40 25 at > 50%F 3.85 3.85 30 at < 50%F 3.00 3.00 25 at < 50%F 3.47 3.47 100 1.46 1.47 1.48 75 1.58 1.58 1.61 65 1.77 1.78 1.79 50
-r3VOOS 50 1.87 1.87 1.94 PLUOOS 40 2.07 2.07 2.16 30 2.36 2.36 2.47 30 at > 50%F 3.26 3.26 3.40 25 at > 50%F 3.70 3.70 3.85 30 at < 50%F 2.85 2.85 3.00 25 at < 50%F 3.29 3.29 3.47 100 1.45 1.45 75 1.58 1.58 65 1.77 1.78 50 FHOOS 50 1.93 1.93 PiUOOS 40 2.16 2.16 30 2.47 2.47 30 at > 50%F 2.90 2.90 25 at > 50%F 3.22 3.22 30 at 5 50%F 2.82 2.82 25 at 5 50%F 3.10 3.10 100 1.48 1.48 75 1.61 1.61 65 1.77 1.78 50 TBVOOS 50 1.94 1.94 FHOOS 40 2.16 2.16 PLUOOS 30 2.47 2.47 30 at > 50%F 3.40 3.40 25 at > 50%F 3.85 3.85 30 at < 50%F 3.00 3.00 25 at S 50%F 3.47 3.47
- All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 2 Cycle 17 Core Operating Limits Report (105% OLTP)
Page 21 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 EIMNPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 4.5 Startup Operation MCPRp Limits for Table 3.1 Temperature Range 1:
Technical Specification Scram Time Basis BOC BOC BOC Operating Pow er to to to End of Condition
(% of rated)
NEOC EOC Coast 100 1.45 1.45 1.45 75 1.58 1.58 1.58 65 1.72 1.72 1.72 50 1.93 1.93 1.93 50 2.11 2.11 2.11 TBV IS 40 2.38 2.38 2.38 30 2.76 2.76 2.76 30 at > 50%F 3.18 3.18 3.18 25 at > 50%F 3.57 3.57 3.57 30 at s 50%F 3.07 3.07 3.07 25 at < 50%F 3.44 3.44 3.44 100 1.48 1.48 1.48 75 1.61 1.61 1.61 65 1.73 1.73 1.73 50 1.94 1.94 1.94 50 2.11 2.11 2.11 TBVOOS 40 2.38 2.38 2.38 30 2.76 2.76 2.76 30 at > 50%F 3.64 3.64 3.64 25 at> 50%F 4.12 4.12 4.12 30 at S 50%F 3.23 3.23 3.23 25 at < 50%F 3.76 3.76 3.76
- Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
Limits are applicable for all non-PLUOOS EOOS scenarios. PLU is inoperable for powers less than 50% rated power, therefore at these powers it can be considered a base case.
Browns Ferry Unit 2 Cycle 17 Core Operadng Limit-. Report, (105% OLTP)
Page 22 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 EIM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 4.6 Startup Operation MCPRp Limits for Table 3.1 Temperature Range 2:
Technical Specification Scram Time Basis*
BOC BOC BOC Pbwer to to to End of Operating Condition
(% of rated)
NEOC EOC Coast 100 1.45 1.45 1.45 75 1.58 1.58 1.58 65 1.72 1.72 1.72 50 1.93 1.93 1.93 50 2.12 2.12 2.12 TBVIS 40 2.40 2.40 2.40 30 2.78 2.78 2.78 30 at > 50%F 3.19 3.19 3.19 25 at > 50%F 3.60 3.60 3.60 30 at* 50%F 3.11 3.11 3.11 25 at 5 50%F 3.46 3.46 3.46 100 1.48 1.48 1.48 75 1.61 1.61 1.61 65 1.73 1.73 1.73 50 1.94 1.94 1.94 50 2.12 2.12 2.12 TBVOOS 40 2.40 2.40 2.40 30 2.78 2.78 2.78 30 at > 50%F 3.65 3.65 3.65 25 at > 50%F 4.13 4.13 4.13 30 at 5 50%F 3.24 3.24 3.24 25 at < 50%F 3.78 3.78 3.78 Limits support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRp limits will be 0.02 higher.
Limits are applicable for all non-PLUOOS EOOS scenarios. PLU is inoperable for powers less than 50% rated power, therefore at these powers it can be considered a base case.
Browns Feny Unit 2 cycle 17 Core Operating Umits Report, (105% OLTP)
Page 23 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 5
APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 24 & 25, and is defined by the following:
SRB _ (0.66(W-AW) + 61%)
Allowable Value SRB _ (0.66(W-AW) + 59%)
Nominal Trip Setpoint (NTSP) where:
SRB
=
Rod Block setting in percent of rated thermal power (3458 MWt)
W
=
Loop recirculation flow rate in percent of rated AW
=
Difference between two-loop and single-loop effective recirculation flow at the same core flow (AW=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
Brcmns Feny unt 2 Cyde 17 Page 24 Browns Ferry Unit 2 Cycle 17 Core Operating Limbt Report, (105% OLTP)
Page 24 TVA-COLR-BF2C17, Re\\Asion 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011
[M NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 6
Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 24 & 25, are shown in Table 6.1. Setpoints are based on an HTSP, unfiltered analytical limit of 114%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superceding values reported in References 24, 25, and 27.
Table 6.1 Analytical RBM Trip Setpoints*
RBM Trip Setpoint LPSP IPSP HPSP LTSP - unfiltered
-filtered ITSP
- unfiltered
- filtered Allowable Value (AV) 27%
62%
82%
121.7%
120.7%
116.7%
115.7%
Nominal Trip Setpoint (NTSP) 25%
60%
80%
120.0%
119.0%
115.0%
114.0%
HTSP -unfiltered 111.7%
110.0%
- filtered 110.9%
109.2%
DTSP 90%
92%
As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 6.2. Cycle specific setpoint analysis results are shown in Table 6.3, per Reference 1.
Table 6.2 RBM Setpoint Applicability Thermal Power Applicable Notes from
(% Rated)
MCPRt Table 3.3.2.1-1 Comment
> 27% and < 90%
< 1.72 (a), (b), (f), (h) two loop operation
< 1.75 (a), (b), (f), (h) single loop operation
> 90%
< 1.42 (g) two loop operation*
Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
t MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
t Greater than 90% rated power is not attainable in single loop operation.
Browns Ferry Unit 2 Cyde 17 Core Operating Limit Report, (105% OLTP)
Page 25 TVA-COLR-BF2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 EIU iNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 6.3 Control Rod Withdrawal Error Results RBM CRWE HTSP Analytical Limit OLMCPR Unfiltered 107 1.31 111 1.35 114 1.36 117 1.39 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., 114% unblocked).
Browns Ferry Unit 2 Cyde 17 Core Operating Limits Report (105% OLTP)
Page 26 TVA-COLR-BF2C17, Revsion 1 (Final)
EDMS: L32 110906 800 IM NPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 7 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
> 0.38% dk/k Browns FerryUnft2 Cyde 17 CoreOperating UmitsReport, (105% OLTP)
Page 27 TVA-COLR-BF2C17, Revision 1 (Final)
EIUNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Appendix A: Thermal-Hydraulic Stability EDMS: L32 110906 800 Date: September 12, 2011 Brc~ns Feny Unit 2 C~de 17 Page 28 Browns Ferry Units2 Cyde 17 Core Operab~ng Limits Report, (105% OLTP)
Page 28 TVA-COLR-BF2C17, R&Asion 1 (Final)
EDMS: L32 110906 800 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 I N PG 1101 Market Street, Chattanooga TN 37402 RPS Instrumentation (Technical Specification 3.3.1.1)
Technical Specification Section 3.3.1.1, LCO 3.3.1.1 states:
The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
Table 3.3.1.1-1, Function 2f, identifies the OPRM upscale function. This function must be operable in conjunction with the following surveillance requirements:
SR 3.3.1.1.1 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17
Background
Browns Ferry uses the Option III stability Detect and Suppress solution as part of the PRNM system. The Option III system is based upon combining groups of local LPRM's into cells known as OPRM's. The OPRM's generate a combined LPRM signal that is examined for the characteristics of a reactor instability event, and if detected, a reactor trip is generated.
The PBDA is the licensing basis portion of the Option III system, requiring a cycle-specific calculation to determine the amplitude setpoint to generate a reactor trip in time to protect the fuel from exceeding the SLMCPR.
The OPRM Upscale Trip function is required to be operable when the plant is in a region of power-flow operation where actual thermal-hydraulic oscillations might occur (T.S. enabled region --
greater than 25% rated thermal power and less than 60% recirculation drive flow).
Setpoints Instrument setpoints are established such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. An Option III stability analysis is performed for each reload core to determine allowable OLMCPR's as a function of OPRM setpoint. Analyses consider both steady state startup operation, and the case of a two recirculation pump trip from rated power.
The resulting stability based OLMCPR's are reported in Reference 1. The OPRM setpoint (sometimes referred to as the Amplitude Trip, Sp) is selected such that required margin to the SLMCPR is provided without stability being a limiting event. Analyses are based on cycle specific DIVOM analyses performed per Reference 23. The calculated OLMCPR's are shown in Blt~nsFenyunt2cx4ie17 Page 29 Browns Ferry Unit 2 Cycle 17 Core operating Limits Report, (105% OLTP)
Page 29 TVA-C0LR-13F2C17, Revision 1 (Final)
EDMS: L32 110906 800 Date: September 12, 2011 IM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table A.1. Review of results, relative to the base case operation shown in COLR Table 4.2 indicates an OPRM setpoint of 1.14 can be supported. Extrapolation beyond a setpoint of 1.15 is not allowed.
Table A.1 OPRM Setpoints OPRM OLMCPR OLMCPR
-Setpoint (SS)
(21PT) 1.05 1.17 1.19 1.06 1.19 1.21 1.07 1.21 1.22 1.08 1.23 1.24 1.09 1.25 1.26 1.10 1.27 1.28 1.11 1.29 1.30 1.12 1.31 1.33 1.13 1.33 1.35 1.14 1.35 1.37 1.15 1.37 1.39 Backup Stability Should the Option III system be declared inoperable, alternate methods/procedures (i.e.,
stability ICA's) are incorporated restricting plant operation in the high power, low core flow region of the power/flow map. ICA's contain specific operator actions, providing clear instructions (depending upon the plant type) for operator response to a reactor inadvertently (or under controlled conditions) entering any of the defined regions. ICA's provide appropriate guidance to reduce the likelihood of hydraulic instability, and enhance early detection in the very unlikely event a stability threshold is exceeded in spite of the ICA guidelines.
In July 2002, GE recommended the original ICAs, established generically in 1994, be re-evaluated to assure adequate conservatism, given the trend to higher energy cores and more aggressive fuel management strategies. The recommended replacement regions and the associated calculational procedure are referred to as BSP, and need to be confirmed on a plant/cycle specific basis. The vendor has performed an ICA/BSP confirmation calculation using the NRC approved method in Reference 22.
Based upon the above discussion, appropriate stability analyses and evaluations have been performed to satisfy licensing requirements.
Browns Ferry Unit 2 cycle 17 Core Operating ULmits Report (105% OLTP)
Page 30 TVA-COLR-BF2C17, Revision 1 (Final)
Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 3 Cycle 15, Core Operating Limits Report, (105% OLTP),
TVA-COLR-BF3CI5, Revision 2 (Final)
(See Attached)
EDMS: L32 110906 801 QA Document Pages Affected: All BFE-2897, Revision 2 N Nuclear Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga, TN 37402 Browns Ferry Unit 3 Cycle 15 Core Operating Limits Report, (105% OLTP)
TVA-COLR-BF3CI5 Revision 2 (Final)
(Revision Log, Page v)
September 2011 Prepared:
/-r -
C/. zli T. W.. Eichenberg, Sr. SpecAist Date:
Date:
q'
,3 Verified:
Approved:
B. C. Nit'hell, Engineer Date:
BWR Fuel Engineering G. C. Storey, Reviewed:
Date:
/
-/-/
W. R. Hayes, ManageroReactor Engineering Approved:
\\
Date:.."
Chairman, PORC U
EDMS: L32 110906 801 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 N PG 1101 Market Street, Chattanooga TN 37402 Table of Contents Total Number of Pages = 33 (including review cover sheet)
L ist o f T a b le s....................................................................................................
iii L ist o f F ig u re s................................................................................................................................................
iv Revision Log...................................................................................................................................
v N o m e n cla tu re..................................................
v i R e fe re n c e s..................................................................................................................................................
v iii I
Introduction..........................................................................................................................
1 1.1 Purpose........................................................................................................................
1 1.2 S c o p e............................................................................................................................
1 1.3 Fuel Loading.................................................................................................................
1 1.4 Acceptability...........................................................................................................
1 2
APLHG R Lim its..........................................................................
3 2.1 Rated Power and Flow Lim it: APLHG RRATED............................................................
3 2.2 Off-Rated Power Dependent Lim it: APLHG Rp.........................................................
3 2.3 Off-Rated Flow Dependent Lim it: APLHG RF............................................................
3 2.4 Single Loop O peration Lim it: APLHG RSLO..............................................................
3 2.5 Equipm ent O ut-Of-Service Corrections.....................................................................
5 3
LHG R Lim its.........................................................................................................................
6 3.1 Rated Power and Flow Lim it: LHG RRATED.................................................................
6 3.2 Off-Rated Power Dependent Lim it: LHG Rp..............................................................
6 3.3 Off-Rated Flow Dependent Lim it: LHG RF......................................................................
6 3.4 Equipm ent O ut-Of-Service Corrections.....................................................................
6 4
O LM CPR Lim its.................................................................................................................
10 4.1 Flow Dependent M CPR Lim it: M CPRF...................................................................
10 4.2 Power Dependent M CPR Lim it: M CPRp................................................................
10 4.2.1 Scram Speed Dependent Lim its (TSSS vs. NSS)...........................................
10 4.2.2 Exposure Dependent Lim its............................................................................
12 4.2.3 Equipm ent O ut-Of-Service (EO OS) O ptions..................................................
13 4.2.4 Single-Loop-O peration (SLO ) Lim its..............................................................
13 4.2.5 Below Pbypass Lim its.....................................................................................
13 5
APRM Flow Biased Rod Block Trip Settings.................................................................
18 6
Rod Block Monitor (RBM) Trip Setpoints and Operability..............
19 7
Shutdown M argin Lim it.................................................................................................
21 Appendix A:
Therm al-Hydraulic Stability...........................................................................
22 Browns Ferry Unit 3 Cyde 15 Page ii Core Operating Limits Report, (105% OLTP)
TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 0I iNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 List of Tables Nuclear Fuel Types..............................................................................................................
2 Nom inal Scram Tim e Basis..................................................................................................
12 M CPRp Lim its for Nom inal Scram Tim e Basis.......................................................................
14 M CPRp Lim its for Technical Specification Scram Tim e Basis...............................................
16 Analytical RBM Trip Setpoints................................................................................................
19 RBM Setpoint Applicability......................................................................................................
19 Control Rod W ithdrawal Error Results...................................................................................
20 O PRM Setpoints.........................................................................................................................
24 Bmwns Feny Unit 3 Cyde 15 Page iii Browns Ferry Unit 3 CyR1e 15 Core Operadng Limits Report (105% OLTP)
Page iii 7VA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801
[U NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 List of Figures APLHGRRATED for ATRIUM-1 0 Fuel..........................................................................................
4 LHGRRATED for ATRIUM-10 Fuel...............................................................................................
7 LHGRFACp for ATRIUM-10 Fuel...........................................................................................
8 LHGRFACF for ATRIUM-10 Fuel............................................................................................
9 MCPRF for ATRIUM-10 Fuel..................................................................................................
11 Browns Ferry Unit 3 Cyde 15 Core Operatng Limits Report, (105% OLTP)
Page iv TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 R ?iNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Revision Log Number Page Description 1-R2 8
Editorial fix. Correct X-axis label to read: Core Power (% Rated)
Revised to support all modes of operation. Converted format to BWRFE report style.
2-R1 vi Eliminated unnecessary Nomenclature items.
Added 5 new references (1-5). Added new methodology references 3-R1 viii-x (22-23) in support of Appendix for OPRM setpoints. Added PRNM setpoint references (24-27).
4-R1 1
Updated Section 1.2 scope to support all modes.
Added new material for Sections 2 through 6. The previous Section 2 5-R1 3-24 becomes new Section 7. Added new appendix discussing OPRM setpoints for RPS instrumentation.
I -RO All New document, per NFTP-1 11, Section 3.3, Item Q.
Brows Fey Uit3 yde15 PgeI Browns Ferrn Unit 3 CyR1e 15 Core Operating Limits Report (105% OLTP)
Page v 1VA-COLR-BF3C1 5, Revision 2 (Final)
EDMS: L32 110906 801 EU NPG.
Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Nomenclature APLHGR APRM AREVA NP BOC BSP BWR CAVEX CD CMSS COLR CPR CRWE CSDM DIVOM EOC EOOS FFTR FFWTR FHOOS ft GWd HTSP ICA ICF IS kW LCO LFWH LHGRFAC LPRM LRNB MAPFAC MCPR MSRV*
Average Planar LHGR Average Power Range Monitor Vendor (Framatome, Siemens)
Beginning of Cycle Backup Stability Protection Boiling Water Reactor Core Average Exposure Coast Down Core Monitoring System Software Core Operating Limits Report Critical Power Ratio Control Rod Withdrawal Error Cold SDM Delta CPR over Initial CPR vs. Oscillation Magnitude End of Cycle Equipment OOS Final Feedwater Temperature Reduction Final Feedwater Temperature Reduction Feedwater Heaters OOS Foot: english unit of measure for length Giga Watt Day High TSP Interim Corrective Action Increased Core Flow (beyond rated)
In-Service kilo watt: SI unit of measure for power.
License Condition of Operation Loss of Feedwater Heating LHGR Multiplier (Power or Flow dependent)
Low Power Range Monitor Generator Load Reject, No Bypass MAPLHGR multiplier (Power or Flow dependent)
Minimum CPR Moisture Separator Reheater Valve Browns Ferry Unit 3 Cycle 15 Core Operating Urmits Report, (105% OLTP)
Page vi TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering - BWRFE Dat M PG 1101 Market Street, Chattanooga TN 37402 e: September 12, 2011 MSRVOOS MSRV OOS MTU Metric Ton Uranium MWd/MTU Mega Watt Day per Metric Ton Uranium NEOC Near EOC NRC United States Nuclear Regulatory Commission NSS Nominal Scram Speed NTSP Nominal TSP OLMCPR MCPR Operating Limit OOS Out-Of-Service OPRM Oscillation Power Range Monitor PBDA Period Based Detection Algorithm Pbypass Power, below which TSV Position and TCV Fast Closure Scrams are Bypassed PLU Power Load Unbalance PLUOOS PLU OOS PRNM Power Range Neutron Monitor RBM Rod Block Monitor RPS Reactor Protection System RPT Recirculation Pump Trip RPTOOS RPT OOS SDM Shutdown Margin SLMCPR MCPR Safety Limit SLO Single Loop Operation TBV Turbine Bypass Valve TBVIS TBV IS TBVOOS Turbine Bypass Valves OOS TIP Transversing In-core Probe TIPOOS TIP OOS TLO Two Loop Operation TSP Trip Setpoint TSSS Technical Specification Scram Speed TVA Tennessee Valley Authority Browns Feny unt 3 Cycle 15 Page vii Browns Ferry Unit 3 Cyde 15 Core Operating Limits Report, (105% OLTP)
Page vii TVA-d0LR-l3F3C15, %Asion 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 References
- 1.
ANP-2895, Revision 0, Browns Ferry Unit 3 Cycle 15 Reload Safety Analysis, AREVA NP, Inc., February 2010.
- 2.
EMF-3213(P) Revision 0, Mechanical Design Report for Browns Ferry Unit 3 Reload BFE3-13 ATRIUM-10 Fuel Assemblies, Framatome ANP, Inc., September 2005.
- 3.
ANP-2628(P) Revision 0, Mechanical Design Report for Browns Ferry Unit 3 Reload BFE3-14 ATRIUMTM -10 Fuel Assemblies, AREVA NP, Inc., May 2007.
- 4.
ANP-2838(P) Revision 0, Mechanical Design Report for Browns Ferry Unit 3 Reload BFE3-15 ATRIUM TM -10 Fuel Assemblies, AREVA NP, Inc., August 2009.
- 5.
ANP-2806(P) Revision 0, Browns Ferry Unit 3 Cycle 15 Plant Parameters Document, AREVA NP, Inc., June 2009.
- 6.
BFE-2904, Revision 1, "Browns Ferry Unit 3 Reload 14 In-Core Shuffle Verification,"
Calculation File, Tennessee Valley Authority, February 2010.
Methodology References
- 7.
XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984.
- 8.
XN-NF-85-67(P)(A) Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.
- 9.
EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998.
- 10.
ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 11.
XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.
- 12.
XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.
- 13.
EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.
- 14.
XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.
Browns Fery unit 3 cycle 15 Core Operating Limis Report (105% OLTP)
Page viii TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering - BWRFE Date: September 12,2011 NPG 1101 Market Street, Chattanooga TN 37402
- 15.
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, Exxon Nuclear Company, February 1987.
- 16.
ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.
- 17.
ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
- 18.
ANF-1358(P)(A) Revision 1, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 1992.
- 19.
EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, Siemens Power Corporation, September 2009.
- 20.
EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001.
- 21.
EMF-2292(P)(A) Revision 0, ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 22.
EMF-CC-074(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
- 23.
BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, Framatome ANP, Inc., May, 2008.
PRNM Setpoint References
- 24.
Filtered Setpoints - EDE-28-0990 Rev. 3 Supplement E, "PRNM (APRM, RBM, and RFM) Setpoint Calculations [ARTS/MELLL (NUMAC) - Power-Uprate Condition] for Tennessee Valley Authority Browns Ferry Nuclear Plant", October 1997.
- 25.
Unfiltered Setpoints - EDE-28-0990 Rev. 2 Supplement E, "PRNM (APRM, RBM, and RFM) Setpoint Calculations [ARTS/MELLL (NUMAC) - Power-Uprate Condition]
for Tennessee Valley Authority Browns Ferry Nuclear Plant", October 1997.
- 26.
GE Letter LB#: 262-97-133, Browns Ferry Nuclear Plant Rod Block Monitor Setpoint Clarification - GE Proprietary Information, September 12, 1997.
- 27.
NEDC-32433P, Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2, and 3, GE Nuclear Energy, April 1995.
Browns Fery Unit 3 cycle 15 Core opeting Limits Report (105% OLTP)
Page ix TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering -BWRFE Date: September 12, 2011 aU PG 1101 Market Street, Chattanooga TN 37402 1 Introduction In anticipation of cycle startup, it is necessary to describe the expected limits of operation.
1.1 Purpose The primary purpose of this document is to satisfy requirements identified by unit technical specification section 5.6.5. This document may be provided, upon final approval, to the NRC.
1.2 Scope This document will discuss the following areas:
Average Planar Linear Heat Generation Rate (APLHGR) Limit (Technical Specifications 3.2.1 and 3.7.5)
Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)
> Minimum Critical Power Ratio Operating Limit (OLMCPR)
(Technical Specifications 3.2.2, 3.3.4.1, and 3.7.5)
Average Power Range Monitor (APRM) Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
> Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
Shutdown Margin (SDM) Limit (Technical Specification 3.1.1) 1.3 Fuel Loading The core will contain all AREVA NP, Inc., ATRIUM-10 fuel. Nuclear fuel types used in the core loading are shown in Table 1.1. The core shuffle and final loading were explicitly evaluated for BOC cold shutdown margin performance as documented in Reference 6.
1.4 Acceptability Limits discussed in this document were generated based on NRC approved methodologies per References 7 through 23.
Brows Fey~ni~c~e15 PgeI Browns Feyng Unit 3 Cy 1Re 15 Core Operaling Limits Report, (105% OLTP)
Page 1 TVA-C0LR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 In NPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 1.1 Nuclear Fuel Types Nuclear Original Number of Fuel Type Fuel Names Fuel Description Cycle Assemblies (NFT)
(Range)
ATRIUM-10 A10-4171B-14GV80-FCB 13 43 1
FCB001-FCB064 ATRIUM-10 A10-4163B-16GV80-FCB 13 68 2
FCB065-FCB232 ATRIUM-10 A10-4181B-13GV80-FCB 13 64 3
FCB233-FCB296 ATRIUM-10 A10-4218B-15GV80-FCC 14 215 4
FCCO01-FCC216 ATRIUM-10 A10-4218B-13GV80-FCC 14 72 5
FCC219-FCC290 ATRIUM-10 Al0-3831B-15GV80-FCD 15 200 6
FCDO01-FCD200 ATRIUM-10 A10-3403B-9GV80-FCD 15 20 7
FCD257-FCB276 ATRIUM-10 A10-3392B-10GV80-FCD 15 36 8
FCD221-FCB256 ATRIUM-10 A10-4218B-15GV80-FCC 15 2
9 FCC217-FCC218 ATRIUM-10 A10-4218B-13GV80-FCC 15 4
10 FCC307-FCC310 ATRIUM-10 Al 0-3757B-1 0GV80-FCC 15 40 11 FCC335-FCC374
" The table identifies the expected fuel type breakdown in anticipation of final core loading. The final composition of the core depends upon uncertainties during the outage such as discovering a failed fuel bundle, or other bundle damage. Minor core loading changes, due to unforeseen events, will conform to the safety and monitoring requirements identified in this document.
Browns Ferry Unit 3 Cycle 15 Core Operating Limits Report (105% OLTP)
Page 2 "VA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 ELU NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 2 APLHGR Limits (Technical Specifications 3.2.1 & 3.7.5)
The APLHGR limit is determined by adjusting the rated power APLHGR limit for off-rated power, off-rated flow, and SLO conditions. The most limiting of these is then used as follows:
APLHGR limit = MIN ( APLHGRp, APLHGRF, APLHGRSLO) where:
APLHGRp APLHGRF APLHGRsLo off-rated power APLHGR limit off-rated flow APLHGR limit SLO APLHGR limit
[APLHGRRATED
- MAPFACp]
[APLHGRRATED
- MAPFACF]
[APLHGRRATED* SLO Multiplier]
2.1 Rated Power and Flow Limit: APLHGRRATED The rated conditions APLHGR, for all fuel types, is identified in Reference 1 and shown in Figure 2.1.
2.2 Off-Rated Power Dependent Limit: APLHGRp Reference 1, for ATRIUM-1 0 fuel, does not specify a power dependent APLHGR. Therefore, MAPFACp is set to a value of 1.0.
2.3 Off-Rated Flow Dependent Limit: APLHGRF Reference 1, for ATRIUM-1 0 fuel, does not specify a flow dependent APLHGR. Therefore, MAPFACF is set to a value of 1.0.
2.4 Single Loop Operation Limit: APLHGRsLo The single loop operation multiplier for ATRIUM-10 fuel is 0.85, per Reference 1.
Br~ns Ferry unt 3 Cyde 15 Page 3 Browns Ferry Unit 3 CyR(e 15 Core Operating Limits Repo*rt, (105% OLTP)
Page 3 TVA-C0LR-BF3C1 5, Revision 2 (Final)
EDMS: L32 110906 801 0
1iNPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 15 12 q
-J I.
6 3
0 0
20 40 60 PlanarAverage Exposure (GWd/MTU) 80 Planar Avg.
Exposure APLHGR Limit (GWd/MTU)
(kWlft) 0.0 12.5 15.0 12.5 67.0 7.3 Figure 2.1 APLHGRRATED for ATRIUM-10 Fuel Browns Ferry Unit 3 Cyde 15 Core Operafing Lmits Report (105% OLTP)
Page 4 IVA-COLR-BF3C15, Revsion 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 0
~NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 2.5 Equipment Out-Of-Service Corrections The limit shown in Figure 2.1 is applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service RPTOOS TBVOOS PLUOOS FHOOS (or FFWTR)
All equipment In-Service (includes 1 SRVOOS)
EOC-Recirculation Pump Trip Out-Of-Service Turbine Bypass Valve(s) Out-Of-Service Power Load Unbalance Out-Of-Service Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction Single Recirculation Loop Operation (SLO) requires the application of the SLO multipliers to the rated APLHGR limits as described previously.
Browns Ferry Unit 3 Cycle 15 Core Operating Umits Report, (105% OLTP)
Page 5 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 R 1NPG Nuclear Fuel Engineering - BWRFE Date:
September 12, 2011 1101 Market Street, Chattanooga TN 37402 3
LHGR Limits (Technical Specification 3.2.3, 3.3.4.1, & 3.7.5)
The LHGR limit is determined by adjusting the rated power LHGR limit for off-rated power and off-rated flow conditions. The most limiting of these is then used as follows:
LHGR limit = MIN (LHGRp, LHGRF) where:
LHGRp LHGRF off-rated power LHGR limit off-rated flow LHGR limit
[LHGRRATED
- LHGRFACp]
[LHGRRATED
- LHGRFACF]
3.1 Rated Power and Flow Limit: LHGRRATED The rated conditions LHGR, for all fuel types, is identified in Reference 1 and shown in Figure 3.1. The LHGR limit is consistent with References 2, 3, and 4.
3.2 Off-Rated Power Dependent Limit: LHGRp The ATRIUM-10 fuel, LHGR limits are adjusted for off-rated power conditions using the LHGRFACp multiplier provided in Reference 1. The multiplier is split into two sub cases:
turbine bypass valves in and out-of-service. The multipliers are shown in Figure 3.2.
3.3 Off-Rated Flow Dependent Limit: LHGRF The ATRIUM-10 fuel, LHGR limits are adjusted for off-rated flow conditions using the LHGRFACF multiplier provided in Reference 1. The multiplier is shown in Figure 3.3.
3.4 Equipment Out-Of-Service Corrections The limit shown in Figure 3.1 is applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS) options; including combinations of the options.
In-Service RPTOOS TBVOOS PLUOOS FHOOS (or FFWTR)
SLO All equipment In-Service EOC-Recirculation Pump Trip Out-Of-Service Turbine Bypass Valve(s) Out-Of-Service Power Load Unbalance Out-Of-Service Feedwater Heaters Out-Of-Service or Final Feedwater Temperature Reduction Single Loop Operation, One Recirculation Pump Out--Of-Service
. All equipment service conditions assume 1 SRVOOS.
Browns Ferry Unit 3 cycle 15 Core operating Limits Report, (105% OLTP)
Page 6 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 IM NPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 The off-rated power corrections shown in Figure 3.2 are dependent on operation of the Turbine Bypass Valve system. For this reason, separate limits are to be applied for TBVIS or TBVOOS operation. The limits have no dependency on RPTOOS, PLUOOS, FHOOS/FFVVTR, or SLO.
15 12 CD
=
-J 9
6 3
0 0
20 40 Pellet Exposure (GWd/MTU) 60 80 Pellet LHGR Exposure Limit (GWd/MTU)
(kWlft) 0.0 13.4 18.9 13.4 74.4 7.1 Figure 3.1 LHGRRATED for ATRIUM-10 Fuel Browns Feny Unt3C~de 15 Page 7 Browns Ferny Unit 3 Cycle 15 Core Operatdng Limits Report, (105% OLTP)
Page 7 TVA-COLR-l3F3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 EIMNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 ILL
_j 1.10 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 20 30 40 50 60 70 80 90 100 110 Core Power (% Rated)
Turbine Bypass In-Service Core Power LHGRFACp
(% Rated) 100.0 1.00 30.0 0.63 Core Flow > 50% Rated 30.0 0.54 25.0 0.49 Core Flow < 50% Rated 30.0 0.58 25.0 0.54 Turbine Bypass Core Power
(% Rated)..
100.0 30.0 Core Flow >
30.0 25.0 Core Flow _5 30.0 25.0 Out-of-Service LHGRFACp 0.90 0.63 50% Rated 0.44 0.40 50% Rated 0.52 0.46 Figure 3.2 LHGRFACp for ATRIUM-10 Fuel (Independent of other EOOS conditions)
Bn~ns Ferry Unit3 C~ie 15 Page 8 Browns Ferry Unit 3 Cycle 15 Core Operating Limits Report, (105% OLTP)
Page 8 TVA-COLR-l3F3C1 5, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 IM ~NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 1.10 1.05 1.00 U-0.95 0.90 0.85 0.80 30 40 50 60 70 80 90 100 Core Flow (% Rated) 110 Core Flow LHGRFACF
(%Rated) 30.0 0.91 47.8 1
107.0 1
Figure 3.3 LHGRFACF for ATRIUM-10 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Br~iis Feny Unit 3 C~e 15 Page 9 Browns Ferry Unet 3 Cy (Fe 15 Core Operatfing Umirts Report, (105% OLTP)
Page 9 TVA-COLR-BF3CI5, ReAsion 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering - BWRFE Date: September12,2011 1101 Market Street, Chattanooga TN 37402 4 OLMCPR Limits (Technical Specification 3.2.2, 3.3.4.1, & 3.7.5)
OLMCPR is calculated to be the most limiting of the flow or power dependent values OLMCPR limit = MAX ( MCPRF, MCPRp) where:
MCPRF core flow-dependent MCPR limit MCPRp power-dependent MCPR limit 4.1 Flow Dependent MCPR Limit: MCPRF MCPRF limits are dependent upon core flow (% of Rated), and the max core flow limit, (Rated or Increased Core Flow, ICF). MCPRF limits are shown in Figure 4.1, per Reference 1. Limits are valid for all EOOS combinations. No adjustment is required for SLO conditions.
4.2 Power Dependent MCPR Limit: MCPRp MCPRp limits are dependent upon:
" Core Power Level (% of Rated)
" Technical Specification Scram Speed (TSSS) or Nominal Scram Speed (NSS)
" Cycle Operating Exposure (NEOC, EOC, and CD - as defined in this section)
" Equipment Out-Of-Service Options
" Two or Single recirculation Loop Operation (TLO vs. SLO)
The MCPRp limits are provided in the following tables, where each table contains the limits for all fuel types and EOOS options (for a specified scram speed and exposure range). The CMSS determines MCPRP limits, from these tables, based on linear interpolation between the specified powers.
4.2.1 Scram Speed Dependent Limits (TSSS vs. NSS)
MCPRp limits are provided for two different sets of assumed scram speeds. The Technical Specification Scram Speed (TSSS) MCPRp limits are applicable at all times, as long as the scram time surveillance demonstrates the times in Technical Specification Table 3.1.4-1 are met. Nominal Scram Speeds (NSS) may be used, as long as the scram time surveillance demonstrates Table 4.1 times are applicable.*t Reference 1 analysis results are based on information identified in Reference 5.
t Assumption basis is consistent with method used to perform actual timing measurements, (i.e., including pickup/dropout effects).
Bopwns Ferry Unit 3 Cy (e
15 Core Opetating Limits Report, (105% OLTP)
Page 10 TVA-C0LR-BF3C1 5, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011
~IMNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 2.00 1.80 1.60
&k IL
-~
~-
1.40 1.20 1.00 30 40 50 60 70 80 90 100 110 Core Flow (% Rated)
Core Flow MCPRF
(% Rated) 7.
30.0 1.61 78.0 1.28 107.0 1.28 Figure 4.1 MCPRF for ATRIUM-10 Fuel (Values bound all EOOS conditions)
(107.0% maximum core flow line is used to support 105% rated flow operation, ICF)
Bim~ns Feny Unit 3 C~de 15 Page 11 Browns Ferny Unit 3 Cycle 15 Core Operating Limits Report, (105% OLTP)
Page 11 TVA-COLR-BF3C1 5, Revision 2 (Final)
1U NPG EDMS: L32 110906 801 Date: September 12, 2011 Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 4.1 Nominal Scram Time Basis Notch Nominal Position Scram Timing (index)
(seconds) 46 0.42 36 0.98 26 1.60 6
2.90 In demonstrating compliance with nominal scram time basis, surveillance requirements from Technical Specification 3.1.4 apply; accepting the definition of SLOW rods should conform to scram speeds shown in Table 4.1. If conformance is not demonstrated, TSSS based MCPRp limits are applied.
On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms NSS based limits are applicable.
4.2.2 Exposure Dependent Limits Exposures are tracked on a Core Average Exposure basis (CAVEX, not Cycle Exposure).
Higher exposure MCPRp limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure. Per Reference 1, MCPRp limits are provided for the following exposure ranges:
BOC to NEOC BOC to EOC BOC to End of Coast NEOC corresponds to EOC corresponds to End of Coast 29,395 MWd / MTU 32,712 MWd / MTU 34,082 MWd / MTU NEOC refers to a Near EOC exposure point.
The EOC exposure point is not the true End-Of-Cycle exposure. Instead it corresponds to a licensing exposure window exceeding expected end-of-full-power-life.
The End of Coast exposure point represents a licensing exposure point exceeding the expected end-of-cycle exposure including cycle extension options.
Browns Feny Unit 3 Cyce 15 Core Operating Umits Report, (105% OLTP)
Page 12 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 E1U NPG Nuclear Fuel Engineering - BWRFE Dat 1101 Market Street, Chattanooga TN 37402 e: September 12, 2011 4.2.3 Equipment Out-Of-Service (EOOS) Options EOOS options* covered by MCPRp limits are given by the following:
In-Service RPTOOS TBVOOS RPTOOS+TBVOOS PLUOOS PLUOOS+RPTOOS PLUOOS+TBVOOS PLUOOS+TBVOOS+RPTOOS FHOOS (or FFVVTR)
All equipment In-Service EOC-Recirculation Pump Trip Out-Of-Service Turbine Bypass Valve(s) Out-Of-Service Combined RPTOOS and TBVOOS Power Load Unbalance Out-Of-Service Combined PLUOOS and RPTOOS Combined PLUOOS and TBVOOS Combined PLUOOS, RPTOOS, and TBVOOS Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)
For exposure ranges up to NEOC and EOC, additional combinations of MCPRp limits are also provided including FHOOS. The CD exposure range assumes application of FFWTR. FHOOS based MCPRp limits for the CD exposure are redundant because the temperature setdown assumption is identical with FFWTR.
4.2.4 Sin-gle-Loop-Operation (SLO) Limits MCPRp limits are increased by 0.02 to support SLO, per Reference 1.
4.2.5 Below Pbypass Limits Below Pbypass (30% rated power), MCPRp limits depend upon core flow. One set of MCPRp limits applies for core flow is above 50% of rated; a second set applies if the core flow is less than or equal to 50% rated.
All equipment service conditions assume 1 SRVOOS.
I Browns Ferry Unit 3 Cycle 15 Core Operating Limits Report (105% OLTP)
Page 13 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 IM NPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 4.2 MCPRp Limits for Nominal Scram Time Basis*
BOC BOC BOC Power to to to End of Operating Condition
(% of rated)
NEOC EOC Coast 100 1.42 1.42 1.44 75 1.52 1.52 1.55 65 1.57 1.57 1.62 50 1.75 1.75 1.81 50 1.91 1.91 1.91 Base Case 40 2.01 2.01 2.03 30 2.23 2.23 2.33 30 at > 50%F 2.53 2.53 2.62 25 at > 50%F 2.77 2.77 2.89 30 at < 50%F 2.47 2.47 2.55 25 at:< 50%F 2.68 2.68 2.79 100 1.46 1.46 1.47 75 1.59 1.59 1.61 65 1.62 1.62 1.64 50 1.77 1.77 1.83 50 1.91 1.91 1.91 TBVOOS 40 2.01 2.01 2.05 30 2.25 2.25 2.35 30 at > 50%F 3:11 3.11 3.23 25 at > 50%F 3.50 3.50 3.62 30 at 5 50%F 2.76 2.76 2.88 25 at < 50%F 3.15 3.15 3.29 100 1.44 1.44 75 1.55 1.55 65 1.62 1.62 50 1.81 1.81 50 1.91 1.91 FHOOS 40 2.03 2.03 30 2.33 2.33 30 at > 50%F 2.62 2.62 25 at > 50%F 2.89 2.89 30 at ! 50%F 2.55 2.55 25 at s 50%F 2.79 2.79 100 1.42 1.43 1.44 75 1.52 1.52 1.55 65 1.81 1.81 1.81 50 50 1.91 1.91 1.91 PLUOOS 40 2.01 2.01 2.03 30 2.23 2.23 2.33 30 at > 50%F 2.53 2.53 2.62 25 at > 50%F 2.77 2.77 2.89 30 at 5 50%F 2.47 2.47 2.55 25 at < 50%F 2.68 2.68 2.79 All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability windoware developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 3 cycle 15 Core Operating Umits Report, (105% OLTP)
Page 14 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801
[U iNPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 4.2 MCPRp Limits for Nominal Scram Time Basis (continued)*
BOC BOC BOC Operating Power to to to End of Condition
(% of rated)
NEOC EOC Coast 100 1.47 1.47 75 1.60 1.60 65 1.64 1.64 50 1.83 1.83
-mVOOS 50 1.91 1.91 EHOOS 40 2.05 2.05 30 2.35 2.35 30 at > 50%F 3.23 3.23 25 at > 50%F 3.62 3.62 30 at < 50%F 2.88 2.88 25 at < 50%F 3.29 3.29 100 1.46 1.46 1.47 75 1.59 1.59 1.61 65 1.81 1.81 1.81 50
-BVOOS 50 1.91 1.91 1.91 P.UOOS 40 2.01 2.01 2.05 30 2.25 2.25 2.35 30at > 50%F 3.11 3.11 3.23 25 at > 50%F 3.50 3.50 3.62 30 at S 50%F 2.76 2.76 2.88 25 at 5 50%F 3.15 3.15 3.29 100 1.44 1.44 75 1.55 1.55 65 1.81 1.81 50 FHOOS 50 1.91 1.91 PiUOOS 40 2.03 2.03 30 2.33 2.33 30 at > 50%F 2.62 2.62 25 at > 50%F 2.89 2.89 30 at: <50%F 2.55 2.55 25 at < 50%F 2.79 2.79 100 1.47 1.47 75 1.60 1.60 65 1.81 1.81 50 TBVOOS 50 1.91 1.91 FHOOS 40 2.05 2.05 PLUOOS 30 2.35 2.35 30 at > 50%F 3.23 3.23 25 at > 50%F 3.62 3.62 30 at. 50%F 2.88 2.88 25at: <50%F 3.29 3.29
" All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 3 Cycle 15 Core Operating Umits Repot, (105% OLTP)
Page 15 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 IU NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 4.3 MCPRp Limits for Technical Specification Scram Time Basis*
BOC BOC BOC Operating Row er to to to End of Condition
(% of rated)
NEOC EOC Coast 100 1.43 1.47 1.52 75 1.56 1.56 1.58 65 1.59 1.59 1.64 50 1.77 1.77 1.83 50 1.92 1.92 1.92 Base Case 40 2.02 2.02 2.05 30 2.25 2.25 2.35 30 at > 50%F 2.53 2.53 2.62 25 at > 50%F 2.77 2.77 2.89 30 at < 50%F 2.47 2.47 2.55 25 at s 50%F 2.68 2.68 2.79 100 1.48 1.52 1.55 75 1.61 1.61 1.63 65 1.63 1.63 1.67 50 1.79 1.79 1.85 50 1.92 1.92 1.92 TBVOOS 40 2.02 2.02 2.07 30 2.27 2.27 2.37 30 at > 50%F 3.11 3.11 3.23 25 at > 50%F 3.50 3.50 3.62 30 at < 50%F 2.76 2.76 2.88 25 at: <50%F 3.15 3.15 3.29 100 1.45 1.47 75 1.58 1.58 65 1.64 1.64 50 1.83 1.83 50 1.92 1.92 FHtOOS 40 2.05 2.05 30 2.35 2.35 30 at > 50%F 2.62 2.62 25 at > 50%F 2.89 2.89 30 at < 50%F 2.55 2.55 25 at < 50%F 2.79 2.79 100 1.44 1.48 1.53 75 1.56 1.56 1.58 65 1.82 1.82 1.82 50 50 1.92 1.92 1.92 PLUOOS 40 2.02 2.02 2.05 30 2.25 2.25 2.35 30 at > 50%F 2.53 2.53 2.62 25 at > 50%F 2.77 2.77 2.89 30 at < 50%F 2.47 2.47 2.55 25 at < 50%F 2.68 2.68 2.79
- All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TI POOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 3 Cyce 15 Core Operating Limits Report (105% OLTP)
Page 16 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801
[U /iNPG Nuclear Fuel Engineering - BWRFE Date: September 12, 2011 1101 Market Street, Chattanooga TN 37402 Table 4.3 MCPRP Limits for Technical Specification Scram Time Basis (continued)*
BOC BOC BOC iower to to to End of Operating condition
(% of rated)
NEOC EOC Coast 100 1.49 1.52 75 1.62 1.62 65 1.67 1.67 50 1.85 1.85 rBVOOS 50 1.92 1.92 F-tOS 40 2.07 2.07 30 2.37 2.37 30 at > 50%F 3.23 3.23 25 at > 50%F 3.62 3.62 30 at ! 50%F 2.88 2.88 25 at < 50%F 3.29 3.29 100 1.48 1.52 1.55 75 1.61 1.61 1.63 65 1.82 1.82 1.82 50 lEVOOS 50 1.92 1.92 1.92 PUOOS 40 2.02 2.02 2.07 30 2.27 2.27 2.37 30 at > 50%F 3.11 3.11 3.23 25 at > 50%F 3.50 3.50 3.62 30 at s 50%F 2.76 2.76 2.88 25 at 5 50%F 3.15 3.15 3.29 100 1.45 1.48 75 1.58 1.58 65 1.82 1.82 50 FHOOS 50 1.92 1.92 nUOOS 40 2.05 2.05 30 2.35 2.35 30 at > 50%F 2.62 2.62 25 at > 50%F 2.89 2.89 30 at s 50%F 2.55 2.55 25 at s 50%F 2.79 2.79 100 1.49 1.52 75 1.62 1.62 65 1.82 1.82 50 TBVOOS 50 1.92 1.92 FHOOS 40 2.07 2.07 PLUOOS 30 2.37 2.37 30 at > 50%F 3.23 3.23 25 at > 50%F 3.62 3.62 30 at < 50%F 2.88 2.88 25 at s 50%F 3.29 3.29
" All limits, including "Base Case," support RPTOOS operation; operation is supported for any combination of 1 MSRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out-of-service. For single-loop operation, MCPRP limits will be 0.02 higher.
FFWTR and FHOOS assume the same value of temperature drop. Consequently, FHOOS limits are not provided for BOC to End of COAST due to redundancy. Thermal limits for the "BOC to End of COAST" exposure applicability window are developed to conservatively bound FHOOS limits for earlier exposure applicability windows.
A 50% power step change for PLUOOS limits is not supported. When core power is < 50%, the LRNB event is the same with, or without PLUOOS.
Browns Ferry Unit 3 Cyce 15 Core Operating Limits Report, (105% OLTP)
Page 17 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 EL! NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 5 APRM Flow Biased Rod Block Trip Settings (Technical Requirements Manual Section 5.3.1 and Table 3.3.4-1)
The APRM rod block trip setting is based upon References 24 & 25, and is defined by the following:
SRB < (0.66(W-AW) + 61%)
SRB < (0.66(W-AW) + 59%)
Allowable Value Nominal Trip Setpoint (NTSP) where:
SRB
=
Rod Block setting in percent of rated thermal power (3458 MWt)
W
=
Loop recirculation flow rate in percent of rated AW
=
Difference between two-loop and single-loop effective recirculation flow at the same core flow (AW=0.0 for two-loop operation)
The APRM rod block trip setting is clamped at a maximum allowable value of 115%
(corresponding to a NTSP of 113%).
Browns Ferry unt 3 C~ie 15 Page 18 Browns Ferry Unit 3 cycle 15 Core Operafing Limbl Report, (105% OLTP)
Page 18 TVA-COLR-BF3C1 5, ReAsion 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 EM NPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 6 Rod Block Monitor (RBM) Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1)
The RBM trip setpoints and applicable power ranges, based on References 24 & 25, are shown in Table 6.1. Setpoints are based on an HTSP, unfiltered analytical limit of 117%. Unfiltered setpoints are consistent with a nominal RBM filter setting of 0.0 seconds; filtered setpoints are consistent with a nominal RBM filter setting less than 0.5 seconds. Cycle specific CRWE analyses of OLMCPR are documented in Reference 1, superceding values reported in References 24, 25, and 27.
Table 6.1 Analytical RBM Trip Setpoints*
RBM Trip Setpoint LPSP IPSP HPSP LTSP - unfiltered
- filtered ITSP
- unfiltered
- filtered HTSP - unfiltered
- filtered Allowable Value (AV) 27%
62%
82%
124.7%
123.5%
119.7%
118.7%
114.7%
113.7%
90%
Nominal Trip Setpoint (NTSP) 25%
60%
80%
123.0%
121.8%
118.0%
117.0%
113.0%
112.0%
92%
DTSP As a result of cycle specific CRWE analyses, RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable as shown in Table 6.2. Cycle specific setpoint analysis results are shown in Table 6.3, per Reference 1.
Table 6.2 RBM Setpoint Applicability Thermal Power Applicable Notes from
(% Rated)
MCPRt Table 3.3.2.1-1 Comment
> 27% and < 90%
< 1.74 (a), (b), (f), (h) two loop operation
< 1.77 (a), (b), (f), (h) single loop operation
> 90%
< 1.43 (g) two loop operation t
. Values are considered maximums. Using lower values, due to RBM system hardware/software limitations, is conservative, and acceptable.
t MCPR values shown correspond with, (support), SLMPCR values identified in Reference 1.
t Greater than 90% rated power is not attainable in single loop operation.
Browns Ferry Unit 3 Cyde 15 Core Opertng Lmits Report, (105% OLTP)
Page 19 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 EIUNPG Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 Table 6.3 Control Rod Withdrawal Error Results RBM CRWE Setpoint OLMCPR Unfiltered 107 1.29 111 1.32 114 1.35 117 1.35 Results, compared against the base case OLMCPR results of Table 4.2, indicate SLMCPR remains protected for RBM inoperable conditions (i.e., unblocked).
BrtM'ns Feny Un~ 3 C~ie 15 Page 20 Browns Ferry Unit 3 Cy1le 15 Core Operating Umils Report, (105% OLTP)
Page 20 IVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering -BWRFE Date: September 12, 2011 M NPG 1101 Market Street, Chattanooga TN 37402 7 Shutdown Margin Limit (Technical Specification 3.1.1)
Assuming the strongest OPERABLE control blade is fully withdrawn, and all other OPERABLE control blades are fully inserted, the core shall be sub-critical and meet the following minimum shutdown margin:
> 0.38% dk/k Browns Ferry Unt 3 Cycle 15 Core Operatng Umits Report (105% OLTP)
Page 21 TVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Date: September 12, 2011 EIUNPG Nuclear Fuel Engineering -
SBWRFE 1101 Market Street, Chattanooga TN 37402 Appendix A: Thermal-Hydraulic Stability Browns Ferry Unit 3 Cyde 15 Core Operating Umits Report (105% OLTP)
Page 22 rVA-COLR-BF3C15, Revision 2 (Final)
EDMS: L32 110906 801 Nuclear Fuel Engineering - BWRFE Date:
September 12,2011 NPG 1101 Market Street, Chattanooga TN 37402 RPS Instrumentation (Technical Specification 3.3.1.1)
Technical Specification Section 3.3.1.1, LCO 3.3.1.1 states:
The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
Table 3.3.1.1-1, Function 2f, identifies the OPRM upscale function. This function must be operable in conjunction with the following surveillance requirements:
SR 3.3.1.1.1 SR 3.3.1.1.7 SR 3.3.1.1.13 SIR 3.3.1.1.16 SR 3.3.1.1.17
Background
Browns Ferry uses the Option III stability Detect and Suppress solution as part of the PRNM system. The Option III system is based upon combining groups of local LPRM's into cells known as OPRM's. The OPRM's generate a combined LPRM signal that is examined for the characteristics of a reactor instability event, and if detected, a reactor trip is generated.
The PBDA is the licensing basis portion of the Option III system, requiring a cycle-specific calculation to determine the amplitude setpoint to generate a reactor trip in time to protect the fuel from exceeding the SLMCPR.
The OPRM Upscale Trip function is required to be operable when the plant is in a region of power-flow operation where actual thermal-hydraulic oscillations might occur (T.S. enabled region --
greater than 25% rated thermal power and less than 60% recirculation drive flow).
SetDoints Instrument setpoints are established such that the reactor will be tripped before an oscillation can grow to the point where the SLMCPR is exceeded. An Option III stability analysis is performed for each reload core to determine allowable OLMCPR's as a function of OPRM setpoint. Analyses consider both steady state startup operation, and the case of a two recirculation pump trip from rated power.
The resulting stability based OLMCPR's are reported in Reference 1. The OPRM setpoint is selected such that required margin to the SLMCPR is provided without stability being a limiting event. Analyses are based on cycle specific DIVOM analyses performed per Reference 23.
The calculated OLMCPR's are shown in Table A.1. Review of results, relative to the base case Browns Ferry Unit 3 Cyce 15 Core Operating Umits Report (10.5% OLTP)
Page 23 TVA-COLR-BF3C15, Revsion 2 (Final)
IM NPG EDMS: L32 110906 801 Date: September 12, 2011 Nuclear Fuel Engineering - BWRFE 1101 Market Street, Chattanooga TN 37402 operation shown in Table 4.2 indicates that an OPRM setpoint of 1.15 can be supported.
Extrapolation beyond a setpoint of 1.15 is not allowed.
Table A.1 OPRM Setpoints OPRM OLMCPR OLMCPR Setpoint (SS)
(2PT) 1.05 1.18 1.18 1.06 1.20 1.20 1.07 1.22 1.22 1.08 1.24 1.24 1.09 1.26 1.26 1.10 1.28 1.28-1.11 1.30 1.30 1.12 1.32 1.32 1.13 1.34 1.34 1.14 1.36 1.36 1.15 1.39 1.39 Backup Stability Should the Option III system be declared inoperable, alternate methods/procedures (i.e.,
stability ICA's) are incorporated restricting plant operation in the high power, low core flow region of the power/flow map. ICA's contain specific operator actions, providing clear instructions (depending upon the plant type) for operator response to a reactor inadvertently (or under controlled conditions) entering any of the defined regions. ICA's provide appropriate guidance to reduce the likelihood of hydraulic instability, and enhance early detection in the very unlikely event a stability threshold is exceeded in spite of the ICA guidelines.
In July 2002, GE recommended the original ICAs, established generically in 1994, be re-evaluated to assure adequate conservatism, given the trend to higher energy cores and more aggressive fuel management strategies. The recommended replacement regions and the associated calculational procedure are referred to as BSP, and need to be confirmed on a plant/cycle specific basis. The vendor has performed an ICA/BSP confirmation calculation using the NRC approved method in Reference 22.
Based upon the above discussion, appropriate stability analyses and evaluations have been performed to satisfy licensing requirements.
Browns Ferry Unit 3 Cyde 15.
Core Operadng Limits Report. (105% OLTP)
Page 24 TVA-COLR-BF3C15, Revision 2 (Final)