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(December 2008 - 302) OL Exam & Ref Materials
	ML090230245
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Site:	Browns Ferry
Issue date:	01/23/2009
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I HLT 0707 NRC Examination

.References CURVE 5 OW SPRAY INIT LIMIT 500__

400 ACTION I"-REQUIRED G:'" a.:::'E*LU 300 l-Ia 200 I I SAFE I 100.0 10 20 30 40 50 60 OW PRESS (PSIG)*CONSTANT ABOVE 60 PSIG CURVE 2 RHR NPSH LIMITS I I 1p PSIG SAFEt--10 PSIG SAFE*--....;.,.

I I 1""00...5PSIG SAFE*-"--........1 0 PSIG SAFE..'""'""-" 245 235 225 Ci:'215ll.205:i:195...J ll.a:: 185 ll.ll.175;:)(/)165 155 145 500 2500 4500 6500 8500 10500 12000 RHR PUMP FLOW (GPM)*SUPPR CHMBR PRESS

NOTE This table is based on 2 year refuel, cycle and a core"off-load" of:::>;*300 fuel bundles.Table 1 Spent Fuel Pool Heat-up Rate at normal Fuel Pool level COLUMNA COLUMNS COLUMN C COLUMN D Decay Time Days'Rate to 125 Rate 125 to 150 Max Temp degrees I hr degrees I hr X y 0 2.7 2.2 180@90hrs 30 2.1 1.6 168@'100 hrs 180 1.3 0.8 152@'144 hrs 365 1.0 0.8 152@144 hrs 730(2 yr cycle)1.0 0.8'152@144 hrs The information provided above is intended to cover all possible event scenarios.

The heat up rates given are for a starting SFSP temperature of 90 degrees and are for the first hourwithou!

any cooling.They will provide a conservative estimate of the time to reach the given temperature.

4.2 SUbsequent Actions (continued)

Use the following formula to determine time to reach"125°F.Use Column A (#of days since the beginning of the last refueling outage)and B to determine current heatup rate.125°F-Actual fuel pool temp (OF)=TIME (in hours)FOR FUEL POOL TO REACH 125°F.X (heatup rate determined from columns A and B (OF I hr))Use the following formula to determine time to reach 150°F.Use COlumn A (#of days since the beginning of the last refueling outage)and C to determine CUfrent heatup rate Time to reach 125 D F+=TIME (in hours)FOR FUEL POOL TOREACH 150"F (calculated above}Y (heatup rate determined from columns A and C eF 1 hr))

3.4 REACTOR COOLANT SYSTEM (RCS)3.4.8 Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES---------------------------------------

1.Both required RHR shutdown cooling subsystems and recirculation pumps may not be in operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.2.One required RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surveillances.

APPLICABILITY:

MODE 4.ACTIONS-------------------------------------------------------NOTE---------------------------------------------------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION A.One or two required RHR A.1 shutdown cooling subsystems inoperable.

REQUIRED ACTION Verify an alternate method of decay heat removal is available for each inoperable required RHR shutdown cooling subsystem.

COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.No RHR shutdown B.1 Verify reactor coolant 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from cooling subsystem in circulating by an alternate discovery of no operation.

method.reactor coolant circulation AND AND No recirculation pump in Once per operation.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND B.2 Monitor reactor coolant Once per hour temperature and pressure.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months subsystem or recirculation pump is operating.

CURVES RPV SATURATION TEMP r------400 380360 en z 340:::l a::: I-z 320 UJ:E:::l 300 a::: I-en280 a:::<<260 UJ z Q.240:E UJ I-220----ACTION._REQUIRED-//'//I SAFE I/.j*200 o 50 100 150 RPV PRESS (PSI G)200 250*CONSTANT ABOVE 250 PSIG CAUTIONS CAUTION#1*AN ffPVlVl NSTRUMENT MAY eE useD TO DETERMiNE CR TREND lVl ONLY WHEN ITReAOSABOIe THE MNIMUM INDICATED lilt ASSOCIA.TEO WITH THE HiGHesT MAX ow 00 SC mJN TEMP.*IF OW TEMPS, CR SC AREA TEMPS (fABLEAS AP'PUCA8lE, ARE OUTSIDE THE SAFE R£Glotol Ol" CURVe 8, THE ASSOCIATED INSTRUMENT MAY 8E UNREuA8l1:

Due TO 80ILNG IN THE RUN.MINIMUM MAX rNIJ RUN TEMP MAXSC INSTRUMENT RANGE INDICATED (FROM XR-64-50 RUN TEMP lVl OR TI-64-52AB)(FROM TABLE 6)ON SCALE NIA ea.OW15O-145 NIA 151 TO 200 L1.3-58A, 8 Et.ERGeNCY

.140 NIA 201 T0250*155 TO ot6O*130 NIA 251 TO 300-120 NIA 301 TO 350 L1*3003 ON SCALE NIA BElOW 150 L1.3-60+5 NIA 151 TO 200 L1-3-206 NORMAl+15 NIA 201 T0250 OTO ot6O L1*3-253+20 NIA 251 TO 300 L1.3-208A, 8, C, 0+30 NIA 301 TO 350 L1.3-Q2 POST L1.3-62A ACOOENT ON SCALE NIA NIA-268 TO+3.2+10 BELOW 100 NIA+15 10010 150 NIA SHUTDOWN+20 151 TO 200 NIA L1-3-55 R.OODJP+30 201 TO 250 NIA OTO+400+40 251 TO 300 NIA+50 301 TO 350 NIA+65 35110400 NIA

-150" TAF-162"-175"-J W Gj-200"-J 0 W()-225 0 Z-250"-268" 3-L1-3-52&62 CORRECTION CURVES-*162"=TAF (RED LINE)*185"=MSCRWL (GREEN LINE)*200"=MZIRWL (BLUE LINE)*215"=TWO*THIRDS CORE HEIGHT (BLACK LINE)" I'-.....1\..!"" I\.....!""I'I""....j"'a 1 00 200 300 400 500 600 700 800 900 1 000 1100 REACTOR PRESSURE (PSIG)ACTUAL LEVEL-162"-185"-200"-215" PIP-95-64 REV.12 TABLE 6 SECONDARY CONTMT INSTRUMENT RUNS INSTRUMENT SC TEMP ELEMENTS AND LOCATIONS EL621 EL593 EL 565 RWCU HXRM (74-95F)(74-95C AND D)(69-835A THRU D)(69-29F.G, H)L1-3-58A of OF N/A of L1-3-58B of of N/A N/A L1-3-53 of of N/A of L1-3-60 of of N/A N/A L1-3-206 of of N/A of L1-3-253 of of N/A N/A L1-3-52 of of of N/A L1-3-62A of of of N/A L1-3-55 of of N/A N/A L1-3-208A.

B of of N/A of L1-3-208C.

D of of N/A N/A

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r TV A Nuclear Fuel Core Operating Limits Report.QADocument BrowhS Ferry Nuclear Plant.Unit 2 Cycle 15 TvA*cOLR*BF2CI5 Revision O.Page 1"..CORE OPERATING LIMITS REPORT.(COLR).TENNESSEE VALLEY AUTHORITY Nuclear Fuel Division BWR Fuel Engineering Department.

...'..Prepared By: VerifiedRy:'

.Approved By:

Date:Z-I'l-o')

.Earl E.Riley, Sf:EllgjI1eeririg

..BWRFueI Engineering

...'"....,;;:O:::!..*Date:,*?/}'1/t!Z*.

.BryeC.Mitchell, Nuclear Engineer..B WR Fuel Engineering'

-.Date: Greg C.Store,.

BWR Fuel ngineering

.Reviewed By:

/:!:e=':::'

-:.-__.*

1.Supervisor

..:'..*.Browns Ferry Reactor Engineering

.Approved By:Revision 0 (2/21/2007)

Date:;..'"'J,..I-o 7 Pages Affected: All TVA Nuclear Fuel Core Operating Limits Report Revision Date o 2/21/2007 Revision Log Description Initial Release for New Cycle TV A-COLR-BF2C15 Revision 0, Page 2 Affected Pages All Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report 1.INTRODUCTION TV A-COLR-BF2C15 Revision 0, Page 3 This Core Operating Limits Report (COLR)for Browns Ferry Nuclear Plant Unit 2 Cycle 15 is prepared in accordance with the requirements of Browns Ferry Technical Specification 5.6.5.This revision of the COLR supports operation at the current licensed thermal power (CLTP)of 3458 MWt which is 105%of original licensed thermal power (OLTP).The core operating limits presented here were developed using NRC-approved methods (References 2 and 3).One exception to this is an issue with the assumed uncertainty for the GEXL 14 CPR correlation.

The NRC has identified that the correlationlackedtop-peaked axial power shape data in its formulation and in the calculation of the overall correlation uncertainty.

As an interim action, an increased GEXLI4 uncertainty that incorporates a significant penalty has been calculated and applied to the MCPR Safety Limit (SLMCPR)for this cycle.Results from the reload analyses for Browns Ferry Nuclear Plant Unit 2 Cycle 15 are documented in Reference I.The following core operating and Technical Specification limits are included in this report: a.Average Planar Linear Heat Generation Rate (APLHGR)Limit (Technical Specifications 3.2.1 and 3.7.5)b.Linear Heat Generation Rate (LHGR)Limit (Technical Specification 3.2.3, 3.3.4.1, and 3.7.5)c.Minimum Critical Power Ratio Operating Limit (OLMCPR)(Technical Specifications 3.2.2, 3.3.4.1, and 3.7.5)d.Average Power Range Monitor (APRM)Flow Biased Rod Block Trip Setting (Technical Requirements Manual Section 5.3.1 andTable3.3.4-1) e.Rod Block Monitor (RBM)Trip Setpoints and Operability (Technical Specification Table 3.3.2.1-1) f.Shutdown Margin (SDM)Limit (Technical Specification 3.1.1)The Unit 2 Cycle 15 core is composed of AREVA-NP ATRIUMTM-lO and Global Nuclear Fuel GE-14Žassemblies.

Throughout this document these are referred to as AIO and GEI4 with the trademark implied.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C15 Revision 0, Page 4 2.APLHGR LIMIT (TECHNICAL SPECIFICATIONS 3.2.1 AND 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 , APLHGR F , APLHGRs LO)where: APLHGR p APLHGR F APLHGRs LO off-rated power APLHGR limit off-rated flow APLHGR limit SLO APLHGR limit[APLHGR RATED*MAPFAC(P)]

[APLHGR RATED*MAPFAC(F)]

[ALPHGR RATED*SLO_Multiplier]

The off-rated power and flow corrections to the APLHGR limit only apply to the GE 14 fuel in the Browns Ferry Unit 2 Cycle 15 core.For that reason, this multiplier is set to 1.0 as shown below for the Al 0 fuel.Rated Power and Flow Limits: APLHGRRATED The APLHGR limits for full power and flow conditions for each type of fuel as a function of exposure are shown in Figures 1-5.The APLHGR limits provided in the COLR figures for the GE14 assemblies are for the most limiting lattice (excluding natural uranium)at each exposure point.The specific values for each GE14 lattice are given in Reference 4.The ATRIUM-l 0 values are provided in Reference 1.Bundle Type Rated Power APLHGR Limit GEI4-PlODNAB416-16GZ (EDB2600)Figure 1 GEI4-PlODNAB416-16GZ (EDB260l)Figure 2 GEI4-PlODNAB416-18GZ (EDB2627)Figure 3 GEI4-PI0DNAB417-18GZ (EDB2628)Figure 4 AI0-3920B-14GV70 Figure 5 AlO-4227B-15GV80-FBB Figure 5 AI0-4239B-15GV80-FBB Figure 5 Al 0-3552B-l OGV80-FBB Figure 5 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report Off-Rated Power Corrections:

APLHGR p TVA-COLR-BF2C15 Revision 0, Page 5 The APLHGR limits for the GEl4 fuel lattices are adjusted for off-rated power conditions using the ARTS multiplier, MAPFAC(P).

The reduced power multiplier, MAPFAC(P), for the GEl4 fuel is provided in Reference 1.No off-rated power correction is required for the AlO rated APLHGR limits.Product Line MAPFAC(P)GEl4 Figure 6 AIO 1.0 Off-Rated Flow Corrections:

APLHGR F The APLHGR limits for the GEl4 fuel lattices are adjusted for off-ratedflowconditions using the ARTS multiplier, MAPF AC(F).The reduced flow multiplier, MAPF AC(F)is provided in Reference 1.No off-rated flow correction is required for the AIO rated APLHGR limits.Product Line MAPFAC(F)GEl4 Figure 7 AIO 1.0 SLO Corrections:

APLHGRs LO Single Recirculation Loop Operation (SLO)requires that the rated power APLHGR limit (APLHGRrated) be reduced by applying the following multipliers.

The GEl4 multiplier is provided in Reference 5.The AlO multiplier is provided in Reference 1.Product Line SLO Multiplier GEl4 0.90*AIO 0.85*The GEl4 SLO multiplier of 0.90 is the more limiting ofCLTP and EPU values provided in Reference 5.This value bounds operation at CLTP.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report Equipment Out..:Of-Service mOOS)Corrections:

TVA-COLR-BF2C15 Revision 0, Page 6 The rated APLHGR limits in Figures 1-5 are applicable for operation with all equipmentService as well as the following Equipment Out-Of-Service (EOOS)options.This includes combinations of these EOOS options.In-Service RPTOOS TBVOOS PLUOOS FHOOS (or FFTR)All equipment In-Service (includes I 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.

The off-rated power corrections

[MAPF AC(P)]in Figure 6 is dependent upon the operating status of the Turbine Bypass Valve (TBV)system.For this reason, separate limits are supplied in these figures to be applied for TBVIS (in service)or TBVOOS (out of service)operation.

The MAPFAC(P)limits have no dependency on RPTOOS, SLO, FHOOSIFFTR, orPLUOOS.The off-rated flow corrections

[MAPF AC(F)]in Figure 7 bound both equipment In-Service or EOOS operation.

Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2C15 Revision 0, Page 7 (3.LHGR LIMIT (TECHNICAL SPECIFICATION 3.2.3, 3.3.4.1, and 3.7.5)The LHGR limit is detennined 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 (LHGR p , LHGRF)where: LHGRp LHGR F off-rated power LHGR limit off-rated flow LHGR limit[LHGRRATED

LHGRFAC(P)]

[LHGR RATED*LHGRFAC(F)]

(The off-rated power and flow corrections to the LHGR limit only apply to the AIO fuel intheBrowns Ferry Unit 2 Cycle 15 core.For that reason, these multipliers for the GEI4 fuel is set to 1.0, as shown below.Rated Power and Flow Limits: LHGRRATED The LHGR limit is fuel type dependent.

The limits for these types are given below: Fuel Type LHGRLimit GEI4 Figure 8 AIO Figure 9 The AIO LHGR limit is provided in References I and 6.The GEI4 LHGR limit is provided in References I and 7.Off-Rated Power Corrections:

LHGR p The LHGR limits for theAI0 fuel are adjusted for off-rated power conditions using the LHGRF AC(P)multiplier which is provided in Reference I.The LHGRF AC(P)multiplier is dependentonwhether the Turbine Bypass system is in-service (TBVIS)or out-of-service (TBVOOS).No off-rated power correction is required for the GEI4 rated LHGR limits.Product Line LHGRFAC(P)

GEI4 1.0 AIO Figure 10 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report Off-Rated Flow Corrections:

LHGRF TVA-COLR-BF2C15 Revision 0, Page 8 The LHGR limits for the AI0 fuel are adjusted for off-rated flow conditions using the LHGRF AC(F)multiplier which is provided in Reference 1.No off-rated flow correction is required for the GE14 rated LHGR limits.Product Line LHGRFAC(F)

GE14 1.0 AlO Figure 11 Equipment Out-Of-Service mOOS)Corrections:

The rated LHGR limits are applicable for operation with all equipment In-Service as well as the following Equipment Out-Of-Service (EOOS)options.This includes combinations of these EOOS options.In-Service RPTOOS TBVOOS PLUOOS SLO FHOOS (or FFTR)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 Single Recirculation Loop Operation Feedwater Heaters Out-Of-Service (or Final Feedwater Temperature Reduction)

The off-rated power corrections

[LHGRF AC(P)]in Figure 10 are dependent upon operation of the Turbine Bypass Valve system.For this reason, separate limits are supplied in this figure to be applied for TBVIS or TBVOOS operation.

The LHGRF AC(P)limits have no dependency on RPTOOS, PLUOOS, SLO, or FHOOSIFFTR.

The off-rated flow corrections

[LHGRF AC(F)]in Figure 11 bound both equipmentService or EOOS operation.

Browns Ferry Nuclear Plimt Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C15 Revision 0, Page 9 4.OLMCPR (TECHNICAL SPECIFICATIONS 3.2.2, 3.3.4.1, AND 3.7.5)The MCPR Operating Limit (OLMCPR)is calculated to be the most limiting of thedependent MCPR (MCPR F)and power-dependent MCPR (MCPR p).OLMCPR limit=MAX (MCPR F , MCPRp)where: MCPRF MCPR p core flow-dependent MCPR limit power-dependent MCPR limit MCPRF limits are provided in Figure 12.MCPR p limits are provided in Tables 1 through 6.Flow-Dependent MCPR Limits: MCPRF The MCPRF limits are dependent upon:*Core Flow (%of Rated)*Max Core Flow Limit (Rated or Increased Core Flow, ICF)*Fuel Type (GEI4 or A 10)The MCPRF limits are provided in Figure 12.For Unit 2 Cycle 15 the same MCPR F limits apply to both the GE14 and AlO fuel types.These limits are valid for all EOOS combinations.

No adjustment is required to the MCPRF limits for SLO.The MCPR F limits are found in Reference 1.Power-Dependent MCPRLimits:

MCPRp The MCPR p limits are dependent upon:*Core Power Level (%of Rated)*Technical Specification Scram Speed (TSSS)or Nominal Scram Speed (NSS)*Fuel Type (GEI4 or A 10)*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 (Ref.1)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 MCPR p limits are determined from these tables using linear interpolation between the specified powers.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report Exposure Range Scram Speed MCPRp BOCtoNEOC NSS Table 1 TSSS Table 2 BOCtoEOC NSS Table 3 TSSS Table 4 BOC to CD NSS Table 5 TSSS Table 6 a.Scram Speed Dependent Limits (TSSS vs.NSS)TV A-COLR-BF2C15 Revision 0, Page 10 MCPR p 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 that the times in Technical Specification table 3.1.4-1 have been met.Nominal Scram Speeds (NSS)may be used as long as the scram time surveillance demonstrates that the times in the following table are met (Ref.9).Notch Nominal Scram Speed Position (seconds)46 0.42 36 0.98 26 1.60 06 2.90 In demonstrating compliance with this table, the same surveillance requirements from Technical Specification 3.1.4 apply, except that the definition of SLOW rods should.conform to the scram speeds in the table above.If conformance to this table is not demonstrated, TSSS MCPRp limits shall be used.On initial cycle startup, TSSS limits are used until the successful completion of scram timing confirms that NSS limits may be used.b.Fuel Type Dependent Limits Separate MCPR p limits are provided for the GE14 and A10 fuel types.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report c.Exposure Dependent Limits TVA-COLR-BF2CI5 Revision 0, Page 11 Exposures are tracked on a Core Average Exposure basis (not Cycle Exposure).

The higher exposure MCPRp limits are always more limiting and may be used for any Core Average Exposure up to the ending exposure.MCPRp limits are provided for the following exposure ranges (Ref.1): BOCtoNEOC BOCtoEOC BOC to CD NEOC corresponds to EOC corresponds to CD corresponds to 27,788 MWd/MTU 31,075 MWd/MTU 32,274 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 that exceeds expected end-of-full-power-life.

The CD (CoastDown) exposure point represents a licensing exposure point that exceeds the expected end-of-cycle exposure including cycle extension options.d.Equipment Out-Of-Service (EOOS)Options EOOS options included in the MCPR p limits are: In-Service RPTOOS TBVOOS RPTOOS+TBVOOS PLUOOS PLUOOS+RPTOOS PLUOOS+TBVOOS PLUOOS+TBVOOS+RPTOOS FHOOS (or FFTR)All equipment In-Service (includes 1 SRVOOS)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 MCPR p limits are also provided that include FHOOS.The CD exposure range assumes application of FFTR, so the CD based MCPRp limits already include FHOOS.e.Single-Loop-Operation (SLO)Limits The MCPRp limits for SLO are to be increased by 0.02 (Ref.1).Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report f.Below Pbypass Limits TV A-COLR-BF2CI5 Revision 0, Page 12 (Below Pbypass (30%rated power), the MCPRp limits are dependent upon core flow.One set of MCPR p limits applies if the core flow is above 50%of rated with a second setthatapplies if the core flow is less than or equal to 50%rated.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2C15 Revision 0, Page 13 5.APRM FLOW BIASED ROD BLOCK TRIP SETTING (TECHNICAL REQUIREMENTS MANUAL SECTION 5.3.1 AND TABLE 3.3.4-1)The APRM Rod Block trip setting shall be (Ref.10): SRB

+61%)SRB

+59%)where: Allowable Value Nominal Trip Setpoint (NTSP)SRB=Rod Block setting in percent of rated thermal power (3458 MWt)W=Loop recirculation flow rate in percent of ratedW=Difference between two-loop and single-loop effective recirculation flow at the same core flow 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 Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2CI5 Revision 0, Page 14 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 shall be as follows (refs.10&11): RBM Trip Setpoint Allowable Value Nominal Trip Setpoint (AY)(NTSP)LPSP 27%25%IPSP 62%60%HPSP 82%80%L TSP-unfiltered 124.7%123.0%-filtered*123.5%121.8%ITSP-unfiltered 119.7%118.0%-filtered 118.7%117.0%HTSP-unfiltered 114.7%113.0%-filtered 113.7%112.0%DTSP 90%92%(1),(2)(1),(2)(1),(2)Notes: (1)These setpoints are based upon an Analytical Limit HTSP of 117%(w/o filter)which corresponds to a MCPR operating limit of 1.42(AIO/GEI4), as reported in section 5.5 of Reference 1.Unit 2 Cycle 15 has had a cycle specific CRWE analysis performed and the table provided in section 5.5 of Reference 1 supercedes the OLMCPR values of references 10 and 12.(2)The unfiltered setpoints are consistent with a nominal RBMfiltersetting of 0.0 seconds (reference 10.b>>.The filtered setpoints are consistent with a nominal RBM filter setting:s 0.5 seconds (reference 10.a>>.The RBM setpoints in Technical Specification Table 3.3.2.1-1 are applicable when: THERMAL POWER Applicable Notes from (%Rated)MCPR (1)Table 3.3.2.1-1 2: 27%and<90%<1.72 (a), (b), (t), (h)<1.75 (a), (b), (t), (h)2:90%<1.47 (g)dual loop operation single loop operation dual loop operation (2)Notes: (1)The MCPR values shown correspond to a SLMCPR of 1.08 for dual recirculation loop operation and 1.10 for single loop operation.(Ref.1).(2)Greater than 90%rated power is not attainable in single loop operation.

Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report 7.SHUTDOWN MARGIN (SDM)LIMIT (TECHNICAL SPECIFICATION 3.1.1)TVA-COLR-BF2CI5 Revision 0, Page 15 The core shall be sub critical with the following margin with the strongest OPERABLE control rod fully withdrawn and all other OPERABLE control rods fully inserted (Ref.8).SDM0.38%dk/k Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report 8.REFERENCES TVA-COLR-BF2C15 Revision 0, Page 16 1.ANP-2592 Rev.0,"Browns Ferry Unit 2 Cycle 15 Reload Analysis for 105%Original Licensed Thennal Power", dated January 2007.2.Framatome-ANP Analytical Methodology

References:

a)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.b)XN-NF-85-67(P)(A)

Revision 1, Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, September 1986.c)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 CQrporation, February 1998.d)ANF-89-98(P)(A)

Revision 1andSupplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.e)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.f)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.g)EMF-2158(p)(A)

Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4IMICROBURN-B2, Siemens Power Corporation, October 1999.h)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.i)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.j)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.k)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.I)ANF-1358(P)(A)

Revision 1, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Advanced Nuclear Fuels Corporation, September 1992.m)EMF-2209(p)(A)

Revision 2, SPCB Critical Power Correlation, Siemens Power Corporation, September 2003.n)EMF-2245(p)(A)

Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.0)EMF-2361(P)(A)

Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001.p)EMF-2292(P)(A)

Revision 0, ATRlUMTM....1O:

Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.3.Global Nuclear Fuel Analytical Methodology

References:

Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2CI5 Revision 0, Page 17 a)NEDE-24011-P-A-15,"General Electric Standard Application for Reactor Fuel", September 2005.b)NEDE-24011-P-A-15-US,"General Electric Standard Application for Reactor Fuel (Supplement for United States)", September 2005.4.0000-0006-1355-MAPL Rev.0,"Lattice-Dependent MAPLHGR Report for Browns Ferry Unit 2 Reload 12 Cycle 13", February 2003.5.NEDC-32484P Rev.6,"Browns Ferry Nuclear Plant Units 1, 2, and3-LOCA Loss-Of-Coolant Accident Analysis", dated February 2005.6.ANP-2537P Rev.0,"Mechanical Design Report for Browns Ferry Unit 2 Reload BFE2-l5 ATRIUMTM-I0 Fuel Assemblies", dated May 2006.7.GE-NE-LI2-00889-00-01P Rev.0,"GEI4 Fuel Design Cycle-Independent Analyses for Browns Ferry Units 2 and 3", dated January 2002.8.TVA-COLR-BF2C14 Rev.1,"Browns Ferry Nuclear Plant Unit 2, Cycle 14 Core Operating Limits Report (COLR)", dated April 10, 2006.9.EMF-3238(p)

Rev.0,"Browns Ferry Unit 2 Cycle 15 Plant Parameters Document", dated January 2006.10.PRNM Setpoint Calculation:

a)Filtered Setpoints-EDE-28-0990 Rev.3 Supplement E,"PRNM (APRM, RBM, and RFM)Setpoint Calculations

[ARTSIMELLL (NUMAC)-Power-Uprate Condition]

for Tennessee Valley Authority Browns Ferry Nuclear Plant", dated October 1997.b)Unfiltered Setpoints-EDE-28-0990 Rev.2 Supplement E,"PRNM (APRM, RBM, and RFM)Setpoint Calculations

[ARTS/MELLL (NUMAC)-Power-Up rate Condition]

for Tennessee Valley Authority Browns Ferry Nuclear Plant", dated October 1997.11.GE Letter LB#: 262-97-133,"Browns Ferry Nuclear Plant Rod Block Monitor Setpoint Clarification

-GE Proprietary Infonnation", dated September 12, 1997.12.NEDC-32433P,"Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1,2, and 3", dated April 1995.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2CI5 Revision 0, Page 18 Exposure of27,788 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Option (0/0 Rated)100 1.48 1.50 69 1.61 1.64 631.661.70 58 1.75---581.781.79 FHOOS 46 1.90 1.97 30 2.33 2.49 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (:s 50%F)2.60 2.79 25 (:S50%F)2.81 3.03 100 1.48 1.50 69 1.61 1.64 631.661.70 58 1.75---RPTOOS 58.1.78 1.79 FHOOS 46 1.90 1.97 30 2.33 2.49 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (:S50%F)2.602.79 25 (:S50%F)2.81 3.03 100 1.51 1.53 69 1.64 1.66 63 1.68 1.72 58 1.77---TBVOOS 58 1.78 1.81 46 1.92 1.98 FHOOS 30 2.35 2.49 30 (>50%F)3.19 3.31 25 (>50%F)3.62 3.78 30 (:S50%F)2.722.88 25 (:s 50%F)3.07 3.30 100 1.51 1.53 69 1.64 1.66 631.681.72 58 1.77---RPTOOS 58 1.78 1.81 TBVOOS 46 1.92 1.98 FHOOS 30 2.35 2.49 30 (>50%F)3.19 3.31 25 (>50%F)3.62 3.78 30 (:S50%F)2.72 2.88 25 (:S50%F)3.07 3.30 Table 1: MCPR p Limits for BOC to NEOC Exposures-NSS Scram Times (A r bi teA e pp lea e up 0 ore vera MCPR p Limit EOOS Power AIO GE14 ODtion (0/0 Rated)100 1.45 1.46 69 1.58 1.62 63 1.61 1.64 58 1.69 1.73 In-581.781.78 Service 46 1.85 1.89 30 2.22 2.35 30 (>50%F)2.642.79 25 (>50%F)2.89 3.08 30 (:s 50%F)2.51 2.68 25 (:S50%F)2.68 2.89 100 1.45 1.46 69 1.58 1.62 63 1.61 1.64 58 1.69 1.73 58 1.78 1.78 RPTOOS 46 1.85 1.89 30 2.22 2.35 30 (>50%F)2.642.79 25 (>50%F)2.89 3.08 30 (:s 50%F)2.51 2.68 25 (:s 50%F)2.68 2.89 100 1.49 1.50 69 1.62 1.66 63 1.65 1.67 58 1.71 1.75 58 1.78 1.78 TBVOOS 46 1.85 1.90 30 2.23 2.36 30 (>50%F)3.09 3.20 25 (>50%F)3.51 3.64 30 (:S50%F)2.642.79 25 (:S50%F)2.97 3.18 100 1.49 1.50 69 1.62 1.66 63 1.65 1.67 58 1.71 1.75 RPTOOS 581.781.78 46 1.85 1.90 TBVOOS 30 2.23 2.36 30 (>50%F)3.093.20 25 (>50%F)3.51 3.64 30 (:S50%F)2.64 2.79 25 (:S50%F)2.97 3.18 Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2CI5 Revision 0, Page 19 Exposure of 27,788 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Ootion (%Rated)1001.481.50 69 1.61 1.64 63 1.73 1.73 58------FHOOS 58 1.78 1.79 461.901.97 PLUOOS 30 2.33 2.49 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (=:;50%F)2.602.79 25 (=:;50%F)2.81 3.03 1001.481.50 69 1.61 1.64 63 1.73 1.73 58------RPTOOS 581.781.79 FHOOS 461.901.97 PLUOOS 30 2.33 2.49 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (=:;50%F)2.60 2.79 25 (=:;50%F)2.81 3.03 100 1.51 1.53 69 1.64 1.66 631.731.73 58------TBVOOS 58 1.78 1.81 FHOOS 46 1.92 1.98 PLUOOS 30 2.35 2.49 30 (>50%F)3.19 3.31 25 (>50%F)3.62 3.78 30 (=:;50%F)2.72 2.88 25 (=:;50%F)3.07 3.30 100 1.51 1.53 69 1.64 1.66 63 1.73 1.73 RPTOOS 58------TBVOOS 58 1.78 1.81 46 1.92 1.98 FHOOS 302.352.49 PLUOOS 30 (>50%F)3.19 3.31 25 (>50%F)3.62 3.78 30 (=:;50%F)2.72 2.88 25 (=:;50%F)3.07 3.30 pp lea eup 0 ore vera MCPR p Limit EOOS Power AIO GE14 Ootion (%Rated)100 1.45 1.46 69 1.58 1.62 631.731.73 58------58 1.78 1.78 PLUOOS 461.851.89 30 2.22 2.35 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (s50%F)2.51 2.68 25 (=:;50%F)2.682.89 100 1.45 1.46 69 1.58 1.62 631.731.73 58------RPTOOS 58 1.78 1.78 46 1.85 1.89 PLUOOS 30 2.22 2.35 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (=:;50%F)2.51 2.68 25 (=:;50%F)2.68 2.89 100 1.49 1.50 69 1.62 1.66 63 1.73 1.73 58------TBVOOS 581.781.78 46 1.85 1.90 PLUOOS 302.232.36 30 (>50%F)3.09 3.20 25 (>50%F)3.5J 3.64 30 (=:;50%F)2.64 2.79 25 (=:;50%F)2.97 3.18 100 1.49 1.50 69 1.62 1.66 631.731.73 58------RPTOOS 581.781.78 TBVOOS 46 1.85 1.90 PLUOOS 30 2.23 2.36 30 (>50%F)3.09 3.20 25 (>50%F)3.51 3.64 30 (=:;50%F)2.64 2.79 25 (=:;50%F)2.97 3.18 Table 1 (Continued):

MCPRp Limits for BOC to NEOC Exposures-NSS Scram Times (A r hI teA e Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2CI5 Revision 0, Page 20 Exposure of27,788 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.50 1.52 69 1.63 1.66 631.681.72 58 1.77---58 1.79 1.81 FHOOS 461.921.99 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (S50%F)2.60 2.79 25 (S50%F)2.81 3.03 1001.501.52 69 1.63 1.66 63 1.68 1.72 58 1.77---RPTOOS 58 1.79 1.81 FHOOS 46 1.92 1.99 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (S50%F)2.602.79 25 (S50%F)2.81 3.03 100 1.53 1.55 69 1.66 1.68 63 1.70 1.74 58------TBVOOS 58 1.79 1.83 46 1.94 2.00 FHOOS 30 2.37 2.52 30 (>50%F)3.20 3.31 25 (>50%F)3.63 3.78 30 (S50%F)2.72 2.88 25 (S50%F)3.083.30 100 1.53 1.55 69 1.66 1.68 63 1.70 1.74 58------RPTOOS 58 1.79 1.83 TBVOOS 46 1.94 2.00 FHOOS 30 2.37 2.52 30 (>50%F)3.20 3.31 25 (>50%F)3.63 3.78 30 (S50%F)2.72 2.88 25 (S50%F)3.08 3.30 Table 2: MCPR p Limits for BOC to NEOC Exposures-TSSS Scram Times (A r bi teA e pp lea eup 0 ore vera MCPRpLimit EOOS Power AIO GE14 Option (%Rated)1001.471.49 69 1.59 1.65 63 1.63 1.66 58 1.71 1.75 In-58 1.79 1.79 46 1.86 1.91 Service 30 2.23 2.37 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (S50%F)2.51 2.68 25 (S50%F)2.68 2.89 100 1.47 1.49 69 1.59 1.65 63 1.63 1.66 58 1.71 1.75 58 1.79 1.79 RPTOOS 46 1.86 1.91 30 2.23 2.37 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (S50%F)2.51 2.68 25 (S50%F)2.68 2.89 100 1.51 1.52 69 1.64 1.65 63 1.66 1.69 58 1.73 1.78 58 1.79 1.79 TBVOOS 46 1.87 1.93 30 2.25 2.39 30 (>50%F)3.10 3.20 25 (>50%F)3.52 3.64 30 (S50%F)2.65 2.79 25 (S50%F)2.97 3.19 100 1.51 1.52 69 1.64 1.65 63 1.66 1.69 58 1.73 1.78 RPTOOS 58 1.79 1.79 46 1.87 1.93 TBVOOS 30 2.25 2.39 30 (>50%F)3.10 3.20 25 (>50%F)3.52 3.64 30 (S50%F)2.65 2.79 25 (S50%F)2.97 3.19 Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2CI5 Revision 0, Page 21 Exposure of27,788 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Option (0/0 Rated)100 1.50 1.52 69 1.63 1.66 63 1.74 1.74 58------FHOOS 58 1.79 1.81 461.921.99 PLUOOS 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (S50%F)2.60 2.79 25 (S50%F)2.81 3.03 100 1.50 1.52 69 1.63 1.66 63 1.74 1.74 58------RPTOOS 58 1.79 1.81 FHOOS 461.921.99 PLUOOS 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (S50%F)2.602.79 25 (S50%F)2.81 3.03 1001.531.55 69 1.66 1.68 63 1.74 1.74 58------TBVOOS 58 1.79 1.83 FHOOS 46 1.94 2.00 PLUOOS 30 2.37 2.52 30 (>50%F)3.20 3.31 25 (>50%F)3.63 3.78 30 (S50%F)2.72 2.88 25 (S50%F)3.08 3.30 100 1.53 1.55 69 1.66 1.68 631.741.74 RPTOOS 58------TBVOOS 58 1.79 1.83 FHOOS 46 1.94 2.00 302.372.52 PLUOOS 30 (>50%F)3.20 3.31 25 (>50%F)3.63 3.78 30 (S50%F)2.72 2.88 25 (S50%F)3.083.30 pp lea e UP to ore vera MCPR p Limit EOOS Power AIO GE14 Option (0/0 Rated)100 1.47 1.49 69 1.59 1.65 631.741.74 58------581.791.79 PLUOOS 46 1.86 1.91 302.232.37 30 (>50%F)2.642.79 25 (>50%F)2.89 3.08 30 (S50%F)2.51 2.68 25 (S50%F)2.68 2.89 100 1.47 1.49 69 1.59 1.65 63 1.74 1.74 58------RPTOOS 581.791.79 46 1.86 1.91 PLUOOS 302.232.37 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (S50%F)2.51 2.68 25 (S50%F)2.68 2.89 100 1.51 1.52 69 1.64 1.65 63 1.74 1.74 58------TBVOOS 581.791.79 46 1.87 1.93 PLUOOS 30 2.25 2.39 30 (>50%F)3.10 3.20 25 (>50%F)3.52 3.64 30 (S50%F)2.65 2.79 25 (S50%F)2.97 3.19 100 1.51.1.52 69 1.64 1.65 63 1.74 1.74 58------RPTOOS 58 1.79 1.79 TBVOOS 46 1.87 1.93 PLUOOS 30 2.25 2.39 30 (>50%F)3.10 3.20 25 (>50%F)3.52 3.64 30 (S50%F)2.65 2.79 25 (S50%F)2.97 3.19 Table 2 (Continued):

MCPRp Limits for BOC to NEOC Exposures-TSSS Scram Times (A r bi CAe Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2CI5 Revision 0, Page 22 Exposure 0[31,075 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Ootion (%Rated)1001.481.50 69 1.61 1.64 631.661.70 58 1.75---58 1.78 1.79 FHOOS 46 1.90 1.97 30 2.33 2.49 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (::;50%F)2.60 2.79 25 (::;50%F)2.81 3.03 1001.481.50 69 1.61 1.64 631.661.70 58 1.75---RPTOOS 58 1.78 1.79 FHOOS 46 1.90 1.97 30 2.33 2.49 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (::;50%F)2.602.79 25 (::;50%F)2.81 3.03 100 1.51 1.53 69 1.64 1.66 63 1.68 1.72 58 1.77---TBVOOS 58 1.78 1.81 FHOOS 461.921.98 30 2.35 2.49 30 (>50%F)3.19 3.31 25 (>50%F)3.623.78 30 (::;50%F)2.72 2.88 25 (::;50%F)3.07 3.30 100 1.51 1.53 69 1.64 1.66 63 1.68 1.72 58 1.77---RPTOOS 58 1.78 1.81 TBVOOS 46 1.92 1.98 FHOOS 30 2.35 2.49 30 (>50%F)3.19 3.31 25 (>50%F)3.62 3.78 30 (::;50%F)2.722.88 25 (::;50%F)3.073.30 Table 3: MCPR p Limits for DOC to EOC Exposures-NSS Scram Times (A r hI teA e pp lea eup 0 ore vera MCPRpLimit EOOS Power AIO GE14 Ootion (%Rated)100 1.45 1.47 69 1.58 1.62 63 1.61 1.64 58 1.69 1.73 10-581.781.78 Service 46 1.85 1.89 30 2.22 2.35 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (::;50%F)2.51 2.68 25 (::;50%F)2.68 2.89 100 1.45 1.47 69 1.58 1.62 63 1.61 1.64 58 1.69 1.73 581.781.78 RPTOOS 46 1.85 1.89 30 2.22 2.35 30 (>50%F)2.642.79 25 (>50%F)2.89 3.08 30 (::;50%F)2.51 2.68 25 (:::50%F)2.682.89 100 1.49 1.51 69 1.62 1.66 63 1.65 1.67 58 1.71 1.75 581.781.78 TBVOOS 46 1.85 1.90 30 2.23 2.36 30 (>50%F)3.093.20 25 (>50%F)3.51 3.64 30 (::;50%F)2.642.79 25 (::;50%F)2.97 3.18 100 1.49 1.51 691.621.66 63 1.65 1.67 58 1.71 1.75 RPTOOS 581.781.78 46 1.85 1.90 TBVOOS 30 2.23 2.36 30 (>50%F)3.09 3.20 25 (>50%F)3.51 3.64 30 (::;50%F)2.642.79 25 (:::50%F)2.97 3.18 Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C15 Revision 0, Page 23 Exposure 0[31,075 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Option (%Rated)1001.481.50 69 1.61 1.64 63 1.73 1.73 58------FHOOS 58 1.78 1.79 PLUOOS 461.901.97 30 2.33 2.49 30 (>SO%F)2.75 2.92 25 (>SO%F)3.03 3.24 30 (SSO%F)2.60 2.79 25 (SSO%F)2.81 3.03 100 1.48 1.50 69 1.61 1.64 63 1.73 1.73 58------RPTOOS 581.781.79 FHOOS 46 1.90 1.97 PLUOOS 30 2.33 2.49 30 (>SO%F)2.752.92 25 (>SO%F)3.03 3.24 30 (SSO%F)2.60 2.79 25 (SSO%F)2.81 3.03 100 1.51 1.53 69 1.64 1.66 63 1.73 1.73 58------TBVOOS 58 1.78 1.81 FHOOS 46 1.92 1.98 PLUOOS 30 2.35 2.49 30 (>50%F)3.19 3.31 25 (>SO%F)3.62 3.78 30 (SSO%F)2.722.88 25 (SSO%F)3.07 3.30 100 1.51 1.53 69 1.64 1.66 63 1.73 1.73 RPTOOS 58------58 1.78 1.81 TBVOOS 46 1.92 1.98 FHOOS 30 2.35 2.49 PLUOOS 30 (>50%F)3.19 3.31 25 (>SO%F)3.62 3.78 30 (SSO%F)2.72 2.88 25 CSSO%F)3.07 3.30 pp lea e up to ore vera MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.45 1.47 69 1.58 1.62 63 1.73 1.73 58------58 1.78 1.78 PLUOOS 46 1.85 1.89 30 2.22 2.35 30 (>SO%F)2.642.79 25 (>SO%F)2.89 3.08 30 (SSO%F)2.51 2.68 25 (SSO%F)2.68 2.89 100 1.45 1.47 69 1.58 1.62 63 1.73 1.73 58------RPTOOS 58 1.78 1.78 46 1.85 1.89 PLUOOS 30 2.22 2.35 30 (>SO%F)2.642.79 25 (>SO%F)2.89 3.08 30 (SSO%F)2.51 2.68 25 (SSO%F)2.68 2.89 100 1.49 1.51 691.621.66 63 1.73 1.73 58------TBVOOS 581.781.78 46 1.85 1.90 PLUOOS 30 2.23 2.36 30 (>SO%F)3.093.20 25 (>SO%F)3.51 3.64 30 (SSO%F)2.642.79 25 (SSO%F)2.97 3.18 100 1.49 1.51 69 1.62 1.66 631.731.73 58------RPTOOS 58 1.78 1.78 TBVOOS 46 1.85 1.90 PLUOOS 30 2.23 2.36 30 (>SO%F)3.09 3.20 25 (>SO%F)3.51 3.64 30 (SSO%F)2.64 2.79 25 (SSO%F)2.97 3.18 Table 3 (Continued):

MCPRp Limits for DOC to EOC Exposures-NSS Scram Times (A r hI CAe (Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2CI5 Revision 0, Page 24 Exposure 0[31,075 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Ontion (0/0 Rated)100 1.50 1.52 69 1.63 1.66 63 1.68 1.72 58 1.77---58 1.79 1.81 FHOOS 46 1.92 1.99 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (:S50%F)2.60 2.79 25 (:S50%F)2.81 3.03 100 1.50 1.52 69 1.63 1.66 631.681.72 58 1.77---RPTOOS 58 1.79 1.81 FHOOS 46 1.92 1.99 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 (:S50%F)2.60 2.79 25 (:S50%F)2.81 3.03 100 1.53 1.55 69 1.66 1.68 631.701.74 58------TBVOOS 58 1.79 1.83 FHOOS 46 1.94 2.00 302.372.52 30 (>50%F)3.20 3.31 25 (>50%F)3.63 3.78 30 (:S50%F)2.72 2.88 25 (:S50%F)3.083.30 1001.531.55 69 1.66 1.68 63 1.70 1.74 58------RPTOOS 58 1.79 1.83 TBVOOS 46 1.94 2.00 FHOOS 30 2.37 2.52 30 (>50%F)3.20.3.31 25 (>50%F)3.63 3.78 30 (:S50%F)2.72 2.88 25 (:S50%F)3.08 3.30 Table 4: MCPRp Limits for BOC to EOC Exposures-TSSS Scram Times (A r bI CAe pp lea e UP to ore vera MCPRpLimit EOOS Power AIO GE14 Option (0/0 Rated)100 1.47 1.49 69 1.59 1.65 63 1.63 1.66 58 1.71 1.75 In-58 1.79 1.79 Service 46 1.86 1.91 30 2.23 2.37 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (:S50%F)2.51 2.68 25 (:S50%F)2.682.89 100 1.47 1.50 69 1.59 1.65 631.631.66 58 1.71 1.75 58 1.79 1.79 RPTOOS 46 1.86 1.91 30 2.23 2.37 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 (:S50%F)2.51 2.68 25 (:s50%F)2.68 2.89 100 1.51 1.52 69 1.64 1.65 63 1.66 1.69 581.731.78 581.791.79 TBVOOS 46 1.87 1.93 30 2.25 2.39 30 (>50%F)3.10 3.20 25 (>50%F)3.52 3.64 30 (:S50%F)2.65 2.79 25 (:S50%F)2.97 3.19 100 1.52 1.53 69 1.64 1.65 631.661.69 58 1.73 1.78 RPTOOS 58 1.79 1.79 46 1.87 1.93 TBVOOS 30 2.25 2.39 30 (>50%F)3.103.20 25 (>50%F)3.52 3.64 30 (:S50%F)2.65 2.79 25 (:s 50%F)2.97 3.19 Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C15 Revision 0, Page 25 Exposure 0[31,075 MWd/MTU)MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.50 1.52 69 1.63 1.66 63 1.74 1.74 58------FROOS 58 1.79 1.81 46 1.92 1.99 PLUOOS 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 302.602.79 25 2.81 3.03 100 1.50 1.52 691.631.66 63 1.74 1.74 58------RPTOOS 58 1.79 1.81 FROOS 46 1.92 1.99 PLUOOS 30 2.35 2.51 30 (>50%F)2.75 2.92 25 (>50%F)3.03 3.24 30 2.60 2.79 25 2.81 3.03 1001.531.55 69 1.66 1.68 63 1.74 1.74 58------TBVOOS 58 1.79 1.83 FROOS 46 1.94 2.00 PLUOOS 30 2.37 2.52 30 (>50%F)3.20 3.31 25 (>50%F)3.63 3.78 30 (s50%F)2.72 2.88 253.083.30 100 1.53 1.55 691.661.68 63 1.74 1.74 RPTOOS 58------TBVOOS 58 1.79 1.83 FROOS 46 1.94 2.00 30 2.37 2.52 PLUOOS 30 (>50%F)3.20 3.31 25 (>50%F)3.633.78 30 2.72 2.88 25 3.08 3.30 pp lea eup 0 ore vera MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.47 1.49 691.591.65 63 1.74 1.74 58------581.791.79 PLUOOS 46 1.86 1.91 30 2.23 2.37 30 (>50%F)2.642.79 25 (>50%F)2.89 3.08 30 2.51 2.68 25 2.68 2.89 100 1.47 1.50 69 1.59 1.65 63 1.74 1.74 58------RPTOOS 581.791.79 PLUOOS 46 1.86 1.91 30 2.23 2.37 30 (>50%F)2.64 2.79 25 (>50%F)2.89 3.08 30 2.51 2.68 25 2.68 2.89 100 1.51 1.52 69 1.64 1.65 63 1.74 1.74 58------TBVOOS 58 1.79 1.79 PLUOOS 46 1.87 1.93 30 2.25 2.39 30 (>50%F)3.10 3.20 25 (>50%F)3.523.64 30 2.65 2.79 25 2.97 3.19 100 1.52 1.53 69 1.64 1.65 63 1.74 1.74 58------RPTOOS 58 1.79 1.79 TBVOOS 46 1.87 1.93 PLUOOS 30 2.25 2.39 30 (>50%F)3.103.20 25 (>50%F)3.523.64 30 2.65 2.79 25 2.97 3.19 Table 4 (Continued):

MCPRp Limits for BOC to EOC Exposures-TSSS Scram Times (A r hI teA e Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2C15 Revision 0, Page 26 Table 5: MCPR p Limits for DOC to CD Exposures-NSS Scram Times (Applicable up to Core Average Exposure of32,274 MWd/MTU)All Values Include FFTRlFHOOS and Bound Heaters In-Service (MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.48 1.50 69 1.61 1.64 63 1.66 1.70 58 1.75---In-581.781.79 Service 46 1.90 1.97 30 2.33 2.49 30 (>SO%F)2.75 2.92 25 (>SO%F)3.03 3.24 30 (SSO%F)2.60 2.79 25 (SSO%F)2.81 3.03 100 1.48 1.50 69 1.61 1.64 63 1.66 1.70 58 1.75---58 1.78 1.79 RPTOOS 461.901.97 30 2.33 2.49 30 (>SO%F)2.75 2.92 25 (>SO%F)3.03 3.24 30 (SSO%F)2.60 2.79 25 (SSO%F)2.81 3.03 1001.511.53 69 1.64 1.66 631.681.72 58 1.77---58 1.78 1.81 TBVOOS 46 1.92 1.98 30 2.35 2.49 30 (>SO%F)3.19 3.31 25 (>SO%F)3.62 3.78 30 (SSO%F)2.72 2.88 25 (SSO%F)3.073.30 100 1.51 1.53 691.641.66 63 1.68 1.72 58 1.77---RPTOOS 58 1.78 1.81 46 1.92 1.98 TBVOOS 30 2.35 2.49 30 (>SO%F)3.19 3.31 25 (>SO%F)3.62 3.78 30 (SSO%F)2.72 2.88 25 (SSO%F)3.07 3.30 MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.48 1.50 69 1.61 1.64 63 1.73 1.73 58------58 1.78 1.79 PLUOOS 461.901.97 30 2.33 2.49 30 (>SO%F)2.75 2.92 25 (>SO%F)3.03 3.24 30 (SSO%F)2.60 2.79 25 (SSO%F)2.81 3.03 100 1.48 1.50 69 1.61 1.64 631.731.73 58------RPTOOS 58 1.78 1.79 PLUOOS 46 1.90 1.97 30 2.33 2.49 30 (>SO%F)2.75 2.92 25 (>SO%F)3.03 3.24 30 (SSO%F)2.60 2.79 25 (SSO%F)2.81 3.03 100 1.51 1.53 69 1.64 1.66 63 1.73 1.73 58------TBVOOS 58 1.78 1.81 PLUOOS 46 1.92 1.98 30 2.35 2.49 30 (>SO%F)3.19 3.31 25 (>SO%F)3.62 3.78 30 (SSO%F)2.72 2.88 25 (SSO%F)3.07 3.30 100 1.51 1.53 69 1.64 1.66 63 1.73 1.73 58------RPTOOS 58 1.78 1.81 TBVOOS 461.921.98 PLUOOS 30 2.35 2.49 30 (>SO%F)3.19 3.31 25 (>SO%F)3.62 3.78 30 (SSO%F)2.72 2.88 25 (SSO%F)3.07 3.30 Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2CI5 Revision 0, Page 27 Table 6: MCPRp Limits for BOC to CD Exposures-TSSS Scram Times (Applicable up to Core Average Exposure of32,274 MWd/MTU)All Values Include FFTR/FHOOS and Bound Heaters In-Service (MCPRpLimit EOOS Power AIO GE14 Option (%Rated)100 1.50 1.52 69 1.63 1.66 631.681.72 58 1.77---In-58 1.79 1.81 46 1.92 1.99 Service 30 2.35 2.51 30 (>500/0F)2.75 2.92 25 (>500/0F)3.03 3.24 30 (:0 500/0F)2.60 2.79 25 (:o500/0F) 2.81 3.03 100 1.50 1.52 69 1.63 1.66 631.681.72 58 1.77---58 1.79 1.81 RPTOOS 46 1.92 1.99 30 2.35 2.51 30 (>500/0F)2.75 2.92 25 (>500/0F)3.03 3.24 30 (:o500/0F)2.602.79 25 (:o500/0F) 2.81 3.03 1001.531.55 69 1.66 1.68 63 1.70 1.74 58------58 1.79 1.83 TBVOOS 46 1.94 2.00 30 2.37 2.52 30 (>500/0F)3.20 3.31 25 (>500/0F)3.63 3.78 30 (:o500/0F) 2.72 2.88 25 (:o500/0F) 3.08 3.30 100 1.54 1.55 69 1.66 1.68 631.701.74 58------RPTOOS 58 1.79 1.83 46 1.94 2.00 TBVOOS 30 2.37 2.52 30 (>500/0F)3.20 3.31 25 (>500/0F)3.63 3.78 30 (:o500/0F)2.722.88 25 (:o500/0F)3.083.30 MCPR p Limit EOOS Power AIO GE14 Option (%Rated)100 1.50 1.52 69 1.63 1.66 63 1.74 1.74 58------58 1.79 1.81 PLUOOS 46 1.92 1.99 30 2.35 2.51 30 (>500/0F)2.752.92 25 (>500/0F)3.03 3.24 30 (:o500/0F) 2.60 2.79 25 (:o500/0F) 2.81 3.03 100 1.50 1.53 69 1.63 1.66 63 1.74 1.74 58------RPTOOS 58 1.79 1.81 46 1.92 1.99 PLUOOS 30 2.35 2.51 30 (>500/0F)2.75 2.92 25 (>500/0F)3.03 3.24 30 (:o500/0F) 2.60 2.79 25 (:o500/0F) 2.81 3.03 100 1.53 1.55 69 1.66 1.68 63 1.74 1.74 58------TBVOOS 58 1.79 1.83 46 1.94 2.00 PLUOOS 30 2.37 2.52 30 (>500/0F)3.20 3.31 25 (>500/0F)3.63 3.78 30 (:o500/0F) 2.72 2.88 25 (:o500/0F) 3.08 3.30 100 1.54 1.55 69 1.66 1.68 63 1.74 1.74 58------RPTOOS 58 1.79 1.83 TBVOOS 46 1.94 2.00 PLUOOS 30 2.37 2.52 30 (>500/0F)3.20 3.31 25 (>500/0F)3.633.78 30 (:o500/0F) 2.72 2.88 25 (:o50%F)3.08 3.30 Add 0.02 to the above MCPR p limits for SLO.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2C15 Revision 0, Page 28 Figure 1 APLHGR Limits for Bundle Type GEl4-PI0DNAB416-16GZ (GE14 EDB#2600)14.00 13.00 12.00 11.00 10.00 at 9.00!8.007.00::i It: C)6.00::t..J 11-c(5.00 4.00 3.00 2.00 1.00 r UNACCEPTABLE OPERATION l......,..-------.:..,...---.................

r-----I ACCEPT ABLE OPERATION I'\*\0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)Most Limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU)(kw/tt)(GWD/MTU)(kw/tt)(GWD/MTU)(kw/tt)0.00 9.41 8.82 10.56 22.05 10.85 0.22 9.51 9.92 10.65 27.56 10.50 1.10 9.61 11.02 10.74 33.07 10.10 2.20 9.73 12.13 10.85 38.58 9.63 3.31 9.86 13.23 10.85 44.09 9.10 4.41 10.00 14.33 10.86 49.60 8.57 5.51 10.14 15.43 10.88 55.12 8.02 6.61 10.28 16.53 10.91 60.63 6.24 7.72 10.42 18.74 10.94 63.50 4.93 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits are for dual recirculation loop operation.

Single Loop Operation (SLO)adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2C15 Revision 0, Page 29 Figure 2 APLHGR Limits for Bundle Type GE14-PI0DNAB416-16GZ (GE14 EDB#2601)14.00 13.00 12.00 11.00 10.00 it 9.00 l 8.00-E 7.00::i It: C)6.00:E:..J Q.Cl: 5.00 4.00 3.00 2.00 1.00 r UNACCEPTABLE OPERATION'

/,...r--.--..................

-......"'" I ACCEPTABLE OPERATION I'\\0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)Most limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU)(kw/ft)(GWD/MTU)(kw/ft)(GWD/MTU)(kwlft)0.00 9.43 8.82 10.76 22.05 10.88 0.22 9.47 9.92 10.83 27.56 10.50 1.10 9.54 11.02 10.91 33.07 10.10 2.20 9.6712.1310.99 38.58 9.66 3.31 9.8313.2311.03 44.09 9.13 4.41 10.02 14.33 11.02 49.60 8.59 5.51 10.21 15.43 11.02 55.12 8.03 6.61 10.4316.5311.03 60.63 6.38 7.7210.6218.74 11.02 63.82 4.92 These values apply to both Turbine Bypass In-Service and Out-Ot-5ervice.

These values apply to both Recirculation Pump Trip In-Service and Out-Ot-Service.

These limits are tor dual recirculation loop operation.

Single Loop Operation (SLO)adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2C15 Revision 0, Page 30 Figure 3 APLHGR Limits for Bundle Type GEl4-PI0DNAB416-18GZ (GE14 EDB#2627)14.00 13.00 12.00 11.00 10.00 t 9.00 8.007.00:::i 0:: Cl 6.00::J:...J a..<5.00 4.00 3.00 2.00 1.00 I UNACCEPTABLE OPERATION l.....-----V--......................

"'\.I ACCEPT ABLE OPERATION I'\-\0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)Most Limiting Lattice for Each Exposure Point Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU)(kw/tt)(GWD/MTU)(kw/tt)(GWD/MTU)(kw/tt)0.00 9.26 8.82 10.54 22.05 10.85 0.22 9.34 9.92 10.65 27.56 10.49 1.10 9.47 11.02 10.75 33.07 10.09 2.20 9.6212.1310.85 38.58 9.60 3.31 9.77 13.23 10.85 44.09 9.09 4.41 9.93 14.33 10.86 49.60 8.56 5.51 10.0915.4310.88 55.12 8.01 6.61 10.25 16.53 10.91.60.63 6.21 7.72 10.4118.7410.93 63.42 4.93 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-5ervice and Out-Of-5ervice.

These limits are for dual recirculation loop operation.

Single Loop Operation (SLO)adjustments are perfonned as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report TVA-COLR-BF2CI5 Revision 0, Page 31 Figure 4 APLHGR Limits for Bundle Type GEl4-PI0DNAB417-18GZ (GE14 EDB#2628)14.00 13.00 12.00 11.00 10.00 it 9.00!8.007.00:::i 0:: Cl 6.00:I:..J Q.c(5.00 4.00 3.00 2.00 1.00 r UNACCEPTABLE OPERATION 1......r---........r--..........." I ACCEPT ABLE OPERATION I'\'\0.00 0.00 10.00 20.00 30.00 40.00 Average Planar Exposure (GWD/MTU)Most Limiting Lattice for Each Exposure Point 50.00 60.00 Average Planar LHGR Average Planar LHGR Average Planar LHGR Exposure Limit Exposure Limit Exposure Limit (GWD/MTU)(kw/ft)(GWD/MTU)(kw/ft)(GWD/MTU)(kwlft)0.00 9.39 8.82 10.74 22.05 10.88 0.22 9.43 9.92 10.83 27.56 10.50 1.10 9.50 11.02 10.91 33.07 10.10 2.20 9.63 12.13 11.00 38.58 9.62 3.31 9.80 13.23 11.02 44.09 9.12 4.41 9.99 14.33 11.02 49.60 8.58 5.51 10.19 15.43 11.02 55.12 8.02 6.61 10.41 16.53 11.03 60.63 6.35 7.72 10.64 18.74 11.02 63.74 4.92 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits are for dual recirculation loop operation.

Single Loop Operation (SLO)adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report Figure 5 APLHGR Limits for all ATRIUM-IOn!

Fuel (A 10)TV A-COLR-BF2C15 Revision 0, Page 32 14.00 13.00 12.00 11.00 10.00 i 9.00 l 8.00...Os 7.00:J a:: Cl 6.00:I:..J Q.<I: 5.00 4.00 3.00 2.00 1.00-.......I UNACCEPTABLE OPERATION'-.........., I ACCEPTABLE OPERATION I 0.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 Average Planar Exposure (GWD/MTU)Average Planar LHGR Exposure Limit (GWD/MTU)(kw/ft)0.00 12.50 15.00 12.50 67.00 7.30 These values apply to both Turbine Bypass In-Service and Out-Of-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Of-Service.

These limits are for dual recirculation loop operation.

Single Loop Operation (SLO)adjustments are performed as described in Section 2 Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report TV A-COLR-BF2C15 Revision 0, Page 33 Figure 6 GE14 Power Dependent MAPLHGR Multiplier

-MAPFAC(P)NSSfTSSS Insertion Times-All Exposures 1.1 0.9 0.8 6:'0.7 g Q.:i 0.6 0.5 0.4 0.3 Turtline Bypass In-Service (TBVIS).............--.....",."".,Turtline Bypass Out-Of-Service (lBVOOS)l""""Tavls:

<50%Core Row TBVIS:: 50%Core ITavOOS:Core I TavOOS:>50%Core Row 0.2 25 30 35 40 455055 60 65 70 758085 90 95 100 Power (%Rated)Turbine Bypass In-Service Core Power MAPFAC(P)(%rated)100 0.89 30 0.48 Core Flow>50%rated 30 I 0.41 25 I 0.38 Core Flow50%rated 30 I 0.46 25 0.43 Turbine Bypass Out-Or-Service Core Power MAPFAC(P)(%rated)100 0.87 30 0.48 Core Flow>50%rated 30 I 0.38 25 I 0.35 Core Flow50%rated 30 I 0.43 25 0.38 (MAPFAC(P)is not dependent upon any Equipment Out-Of-Service except Turbine Bypass.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TVA Nuclear Fuel Core Operating Limits Report Figure 7 Flow Dependent MAPLHGR Factor-MAPFAC(F)(GE14)TVA-COLR-BF2C15 Revision 0, Page 34 1.00 0.90 it 0.80 U If II.:Ii 0.70 0.60 IV MAX FLOW=102.5%\d V V\MAX FLOW*107%0.50 30 40 50 60 70 80 90 100 Max Core Flow 102.5%Rated Core Flow MAPFAC(F)(%rated)30 0.62 71 1.00 102.5 1.00 Core Flow ("to Rated)Max Core Flow 107%Rated Core Flow MAPFAC(F)(%rated)30 0.60 75 1.00 107 1.00 These values bound both Turbine Bypass In-Service and Out*Of-Service.

These values bound both Recirculation Pump Trip In-Service and Out-Of-Service.

The 102.5%maximum flow line is used for operation up to 100%rated flow.The 107%maximum flow line is used for operation up to 105%rated flow (ICF).Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report Figure 8 LHGR Limits for aU GE-14 Fuel (GE14)TVA-COLR-BF2C15 Revision 0, Page 35 14.00 13.00 12.00 11.00 10.00 9.00!8.00:!:: E 7.00::i a: 6.00 C>:%:..J 5.00 4.00 3.00 2.00 1.00...........r UNACCEPTABLE OPERATION a........................

i"-...................

i"-....\I ACCEPTABLE OPERATION I\\0.00 0.00 10.00 20.00 30.00 40.00 50.00 Pellet Exposure (GWD/MTU)Pellet LHGR Exposure Limit (GWD/MTU)(kw/ft)0.00 13.40 16.00 13.40 63.50 8.00 70.00 5.00 60.00 70.00 These values apply to both Turbine Bypass In-Service and Out-Ot-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Ot-Service.

These limits apply to both Two Loop Operation (TLO)and Single Loop Operation (SLO).Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report Figure 9 LHGR Limits for all ATRIUM-tO Fuel (A 10)TV A-COLR-BF2C15 Revision 0, Page 36 14.00 13.00 12.00 11.00 10.00 9.00 it!8.00-E 7.00::::i 0:: 6.00 C):c...J 5.00 4.00 3.00 2.00 1.00-..........I UNACCEPTABLE OPERATION'........'"............

...............

I ACCEPTABLE OPERATION I 0.00 0.00 10.00 20.00 30.00 40.00 50.00 Pellet Exposure (GWD/MTU)60.00 70.00 Pellet LHGR Exposure Limit (GWD/MTU)(kw/ft)0.00 13.40 18.90 13.40 74.40.7.10 These values apply to both Turbine Bypass In-Service and Out-Or-Service.

These values apply to both Recirculation Pump Trip In-Service and Out-Or-Service.

These limits apply to both Two Loop Operation (TLO)and Single Loop Operation (SLO).Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report Figure 10 AIO Power Dependent LHGR Multiplier

-LHGRF AC(P)NSSfTSSS Insertion Times-All Exposures Revision 0, Page 37 1.10 1.00 0.90 0.80rr if 0.70...0.60 0.50 0.40---------io""".----Turbine Bypass In-Service (TBVIS)I-""""'"'.....................

-r'-..p-Turbine Bypass Out-Ot-Service (TBVOOS)TBVIS:50%Core Row ITBVOOS:50%Core TBVlS:>50%Core VIII/'TBVOOS:>50%Core Row 0.30 253035 40 455055 606570 75 BO B5 90 95 100 Turbine Bypass In-Service Core Power LHGRFAC(P)

(%rated)100 0.93 30 0.64 Core Flow>50%rated 30 0.51 25 0.46 Core Flow50%rated 30 0.55 25 0.50 Power (%Rated)Turbine Bypass Out-Of-Service Core Power LHGRFAC(P)

(%rated)100 0.93 30 0.63 Core Flow>50%rated 30 I 0.45 25 I 0.39 Core Flow50%rated 30 I 0.55 25 I 0.48 LHGRFAC(P) is not dependent upon any Equipment Out-Of-5ervice except Turbine Bypass.Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report Figure 11 Flow Dependent LHGR Multiplier

-LHGRF AC(F)(AIO Fuel)TVA-COLR-BF2C15 Revision 0, Page 38 1.00 0.95 iL UII:: 0.90 Cl:J:....0.85 I V/\/l'i V\MAX FLOW c 107%/0.80 30 40 50 60 70 80 90 100 Max Core Flow 102.5%Rated Core Flow LHGRFAC(F)

(%rated)30 0.91 48 1.00 102.5 1.00 Core Flow (%Rated)Max Core Flow 107%Rated Core Flow LHGRFAC(F)

(%rated)30 0.88 54.4 1.00 107 1.00 These values bound both Turbine Bypass In-Service and Out-Of-Service.

These values bound both Recirculation Pump Trip In-Service and Out-Of-Service.

The 102.5%maximum flow line is used for operation up to 100%rated flow.The 107%maximum flow line is used for operation up to 105%rated flow (ICF).Browns Ferry Nuclear Plant Unit 2 Cycle 15 TV A Nuclear Fuel Core Operating Limits Report Figure 12 Flow Dependent MCPR Limit-MCPR(F)(All Fuel)TV A-COLR-BF2C15 Revision 0, Page 39 (Final)1.43 1.38 1.33it IL (,)==1.28 1.23""MAX FLOW'107%'":<MAX FLOW'102.5%'"............

1.18 30 40 60 70 80 90 100 Max Core Flow 102.5%Rated Core Flow MCPR(F)(%rated)30 1.37 72 1.21 102.5 1.21 Core Flow (%Rated)Max Core Flow 107%Rated Core Flow MCPR(F)(%rated)30 1.40 78 1.21 107 1.21 These values bound both Turbine Bypass In-Service and Out-Of-Service.

These values bound both Recirculation Pump Trip In-Service and Out-Of-Service.

The 102.5%maximum flow line is used for operation up to 100%rated flow.The 107%maximum flow line is used for operation up to 105%rated flow (ICF).Browns Ferry Nuclear Plant Unit 2 Cycle 15

RPg*linstrunlentation 3.3=0'1.1 3.3'a1 0'1'The RiPS i:nstrllBlentationfor FUl1ebon:in Table 3..3.'1:.

stlaU be'O,PEjRABlEy APPL I:CAB tlITY':

T 3=..:1.."1-,:1..

-.-------.----.-------.------------------------.--------*N.O TE------..--.----------.--.------.-.--------------------.------

Se'parate entry' for e:ach channel C:ONDiTtON A..OrJ:e:or nl1)re re:qu:ired cjha:nner.s

in:op'erable

..A..2-*--*----------N:OTE-----------*--

N.ol.applrhcabfefor or associated trip s')tstem:lf1

'trip_

12 hours:12 hours Amendnlel1t:

March (t5: J

.RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.-----NOTE------

B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for system in trip_Functions 2.a, 2.b, 2.c, 2.d, or 2.1.OR----------------

8.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months One or more Functions trip.with one or more required channels inoperable in both trip systems.C.One or more Functions C.1 Restore RPS trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability capability

.not maintained.

D.Required Action and 0..1 Enter the Condition Immediately associated Completion referenced in Time of Condition A.B f or Table3.3.1 i.1-1 for the C not met.channel.E As required by Required E1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to<30%RTP.referenced in Table 3.3.1.1-1.

F.As required by Required F.1 Be in MODE 2.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.l and referenced in Table 3.3.1.1-1.

(contInued)

BFN-UNIT 2 3.3-2 Amendment No.258 March 05, 1999 RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G.As required by Required G.1 Be in MODE 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action 0.1 and referenced in Table 3.3.1.1-t H.As required by Required H.1 Initiate action to fully Immediately Action 0.1 and insert all insertable referenced in control rods in core cells Table 3.3.1.1-1.

containing one or more fuel assemblies.

l.As required by Required 1.1 Initiate alternate method 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action 0.1 and to detect and suppress referenced in Table thermal hydraulic 3-3.1.1-1.

instability oscillations.

J.Required Action and J.1 Be in Mode 2.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time of Condition I not met BFN-UNIT 2 3.3-3 Amendment No.273 July 26, 2001 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS-NOTES------------------

--1.Refer to Table 3.3.1.'1-1 to determine which SRs apply for each RPS Function.2.When a channel is placed in an inoperable status solely for performance of required Sur"eillances, entry into associated Condttions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

(SR 3..3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.3 BFN-UNIT 2 SURVEILLANCE Perform CHANNEL CHECK.-----------NOTE------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER25%RTP.Verify the absolute difference between the average power range monitor (APRM)channels and the calculated power is2%RTP while operating at25%RTP.

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.Perform CHANNEL FUNCTIONAL TEST.3.3-4 FREQUENCY 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7 days 7 days (continued)

Amendment No.253 RPS fnstrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 33.1.1.4 Perform CHANNEL FUNCTIONAL TEST.7 days SR 3.3.1.1.5 Verify the source range monitor (SRM)and Prior to intermediate range monitor (IRM)channels withdrawing overlap.SiRMs from the fully inserted position SR 3.3.1.1.6-------------NOTE-------------

Only requirred to be met during entry into MODE 2 from MODE 1.---------------------------------

Verify the IRM and APRM channels overlap.7 days SR 3.3.1.1.7 Calibrate the local power range monitors.1000 MWDfT average core exposure SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST.92 days SR 3.3.1.1.9--------------------NOTES---------------------

1.Neutron detectors are excluded.2.For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.-------------------------------------

Perform CHANNEL CALIBRATION.

92 days (continued)

BFN-UNIT 2 3.3-5 Amendment No.253 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1:UO Perfonn CHANNEL CALIBRATION.

184 days SR 3.3.1.1.11 (Deleted)SR 33.1.1.12 Perform CHANNEL FUNCTIONAL TEST.24 months SR 3.3.1.1.13

NOTE-------------

Neutron detectors are excluded.----------------------------

Perform CHANNEL CALIBRA TrON.24 months SR3..3.1.1.14 Perfom, LOGIC SYSTEM FUNCTIONAl 24 months TEST.SR 3.3.1.1.15 Verify Turbine Stop VaJve-Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are not bypassed when THERMAL POWER is30%RTP.SR 3..3.1.1.16-------------NOTE-------------

For Function 2.3, not required to be perfonned when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.-----------------------------------

Perform CHANNEL FUNCTIONAL TEST.184 days SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is 2: 25%and recirculation drive flow is<60%of rated recirculation drive flow.(BFN-UNIT 2 3.3-6 Amendment No.258 March 05, 1999 RPS Instrumentation 3.3.1.1 Table 33.1.1-1{page 1 of3}Reactcr plltJec:liDn System IRfilrumerrta'iicn FUNCTlON*Af"?UCABt.E ltlODESOR OTHER SPECIAED CONDITIONS

REQLRRED CHANNElS FBRTRIP SYSTEM CCNDlTIONS REFERENCED FROM REQUIRED ACTICND.1*SURVEILLANCE AlLOINA8LE REOUIREMENTS VAlUE b.Incp 2: 3 G 5l a)3 H 2-Allerage P<Mef Range MooilDn;a.Neutron Aux-High, 2 3{b)G{SetOOt.n}

1.Intamediate Range Moni!D1:5 a.Ne!J1ran AWl-High b.Flew Biaaed Simulated Thelmal Power-High e.Neulroo Aux-High 2 5l a}3 3G H F F SR3..3.1.1.1 SR 3.3.1.1..3 SR 3.3.1.1.53.3.1.1.6 SR 3.3.1.Ul SiR 3.3.1.1.14 SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.9 SR 1.3.1.1.14 SR 3.3.1.1.3 SR 3.3.1.1.14 SR 3.3.1.1.4 SiR 3.3.1.1.14 SiR 3.3.1.1.1 SR 3.3.1.1.6 SR 3..3.1.1.7 SiR 3.3.1.1.13 SR 3.3.1.1.111 SR 3.3.1.1.1 SiR 3.3.1.12 SR 3.3.1.1.7 SR 33.1.1.133.3.1.1.111 SR 3.3.1.1.1 SR 3.3.1.1.2£'R 3.3.1.1.7 SR 3.3.1.1.13 SR:l.3.1.1.16,:;;12Ol'125 dl**4(sicnr.

of flil scale$12Of125 dlYi6ionr.

of flil&::ale:$15%RlP$Q.6!l W+66"t.RT?and:f12D"t.RTF{c):$120%RTF (a)With NPJ oontrcl red from a COle cell containing one or mere fuel as&entlIie6.(h)Each APRM channel JX'lMder.inputr.lD both trip aystEms.{c)[.66 W..00"'-.66 d W]RTF wtJenfl:ningle loop operation per LC03.4.1,"Re::i"cdalion lDcp6 Opsaling.NOTE: This page is applicable after commencing Cycle 11 operation.

BFN-UNIT 2 3.3-7 Amendment No.256 December 23, 1998 RPS I'nstrumentation 3.3.1.1 Table 33.1.1-t{p.age 2m3)Reactcr-PrcfJecIion FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDmONS REQUIRED CHANNElS PER TRIP SYSTEM CONOlTlONS REFERENCED FROM S1.JRVEIUANCE ALLOWABLE REQUIRED REQUIREMENTS VALUE At;<Jt4 0.1 2.Alrerage Pcwer Range Moniter&.(conlinllf!d) d.loop e.2-OLrt.Qf-1 Vater f.OPRM Upscale 3.ReaclXlr Ve&&el steam Dame Pleiwre..Hign{d)4.Reactor VI!6&E!Waled.ewl..1.atI;LM 3(d)5.Main SlPam I&oIaIlcnOCl6Uli!6.Dytlel Preli6ure..High 7.Saam fA6d1i1l'Ql!

Volume Waier leIIel-High 1,2 1,2 1,2 1.2 1.22 2 2 8 2 G G G G F G!OR:U.1.1.16 SR 3.3.1.1.1 SR 3.3.1.1.14 SR 3.3.1.1.16

!OR 3.3.1.1.1!OR 3.3.1.1.7 SR 3.3.U.13 SR 3.3.1.1.16 SR 3.3.U.17 SR 33.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10

!OR 3.3.1.1.14 SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.U.13 SR 33.1.1.14 SR 3.3.1.1.8 SR 33.U.13 SR 33.1.1.14 SR 3.3.U.8 SR 33.1.1.13 SR 3.3.U.14 NA!'fA NAlllOOpsig 510%clO&ed 1.2 2 G SR 3.31.1.8::;00 gaBcm;!OR 3.3.1.1.13 SR 3.3.1.1.14 5<a)2 Hi SR 3.3.1.1.8$50 gaBcm;SR 3.3..1.1.13 SR 33.1.1.14 (4)Vdln arI).c:rIlrct roo wlUllfroNinllllll'll a 00ll!<:ell oonta"iIlrlg CllIl!or II1In Il!Il!I a&UITlbllei.

\11')Ea:h APR!\!cltannEl IJ)beth IlIp syilHni.(0')DuI1llg Initrument 111lle Ai Found d!1arIlel HlpaIlllI6Cll11Se1'IiII'le

'MIll re6peCt1ll t!t£o Nlcwal:ille Value but oulikle Ito aDOepl1ae N.FoonlI band as c1Slrted: by Ito asscx:ta:-!!ll Requlferrsrt proCEdUre.

1tIe.'1l1le!&!

ilIa1lbe an Il'IDilI 1D en6lR oonlllliEflce 1Ilal1lli!J:"".n pEf1"t:ml as IeqUlreclj IleIllR Il!lUTlIng 1Ile c:hamello 6ECV!lle In SlllVl!IIiioOl!.

Ai Found Im'J\IIlEI'li cI!aMEl IS-nat141111 ra;ped III 1M/lJIcWaIll!!

\J3lU!-.1IIe.ctalU!E!:

WIlle llel:faIed POOr 10 rR.mlng a cIIanIlEI11D

&e1\1l:e.jill!1nSr.n.l1lEl't C1lilm!!1&elpillllt

&IIaJ be calJlr.3ied m a lIakJ!!that Ii VlIl!trl1lleM lBllOlerallCe

<<1lIe ie1pal"lt cltK!f\\1W, lhi!*cIJamleI61m be*lle!S<II!!I mperaIlli!."ill!!IICn1l'1aI Tt1p Sle1pclrollillallleon (Il!i6l!P'1 culpUt OOCUrD;!1lt1iilan Wllk:lt 15 mCQ1'jlOlilla:l b)'In theArB sarety Rep<<t The*melr.ocll:lcgy U&e!l1XI ile'.eI1l1Inl!'

the nomIrBSElpafnt.!III!

pll!(Il!IInecl N.FDtm aoo llle A6 LSll llli!1iIOOe' IlilnU,.iIlId a D;Ung Cflhi!6I!lpolnt oolput illlCUmentalJan 6ln1i be ipec:tll2lJ In Cl'liIplS" 7 CIT 1M Final sa:ety:Anal)'&l1>

RiepCrt.BFN-UNIT 2 3.3-8 Amendment No. 264, 2a8, 2GQ, 296 September 14, 2006

(Primary Containment Isolation Instrumentation 3.3.6.1 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation Leo 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.APPLICABILITY:

According to Table 3.3.6.1-1.

BFN-UNIT 2 3.3-53 Amendment No.253 Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS----------------------------------------NOTE---------------------------------------------

Separate Condition entry is allowed for each channel.CONDlTrON A.One or more required channels inoperable.

BFN-UNIT 2 REQUIRED ACTION A.1------NOTE---------

Only applicable tor Function 1.d if h'fO or more channels are inoperable.

Place channel in trip.AND A.2---------NOTE---------

Only applicable for Function 1.d when 15 of 16 channels are OPERABLE.Place channel in trip.3.3-54 COMPLETION TIME'12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Functions 2.a, 2.b, 5.h, G.b, and S.C 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Functions other than Functions 2.a, 2.b, 5.h, G.b, and G.c 30 days (continued)

Amendment No.253 Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.One or more Functions a.-I Restore isolation 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with isolation capability capability.

not maintained.

OR 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Function*I.d when normal ventilation is not available C.R,equired Action and C.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A or B Table 3.3.6.1i-*1 for the not met channel.D.As required by Required 0.1 Isolate associated Main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action C.-r and Steam line (MSL).referenced in Table 3.3.6.1-1.

OR 0.2.1 Be in MODE 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND 0.2.2 Be in MODE 4.36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

BFN-UNIT 2 3.3-55 Amendment No.253 Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDIT'ON REQUIRED ACTION COMPLETION TIME E.As required by Required E.-t Be in MODE 2.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.t and referenced in Table 3.3.6.1-1..

F.As required by Required F.1 Isolate the affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action C.'f and penetration flow path(s).referenced in Table 3.3.6.1-1.

G.As, requir,ed by Required G.1 Be in MODE 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action C.1 and referenced in AND Table 3.3.6.1-1.

G.2 Be in MODE 4.36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Required Action and associated Completion Time for Condition F not met.(continued)

BFN-UNIT 2 3.3-56 Amendment No.253 Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION HME H.As required by Required H.1 Declare standby liquid'1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action C.l and control system (SLC)referenced in inoperable.

Table 3.3.6.1-1.

OR H.2 Isolate the Reactor Water 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Cleanup System.t As required by Required 1.1 Initiate action to restore Immediately Action C.1 and channel to OPERABLE referenced in status.Table 3.3.6.1-1.

OR 1.2 Initiate action to isolate Immediately the Residual Heat Removal (RHR)Shutdown Cool'ing System.BFN-UNIT 2 3.3-57 Amendment No.253 Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS

NOTES-------------------------------------

1.Refer to Table 3.3.6.1-1 to determine ylhich SRs apply for each Primary Containment IsolatEoI1 Function.2.When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Functi,on maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.*U Perform CHANNEL CHECK.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST.92 days SR 3.3.6.1.3 Perform CHANNEL CALIBRATION.

92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION.

122 days SR 3.3.6.1.5 Perform CHANNEL CALIBRATION.

24 months SR 3.3.6o*li.6 Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST.BFN-UNliT 2 3.3-58 Amendment No.255 November 30, 1998 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3..3.6.1-1 Me 1 of 3)PriIrory Ccntaimnfflt IEdation Inslrument\1ioo***".P?UCABt:E CONDITIONS MOOESOR REQURED REFERENCED F1JNCT1CN OTHER CHANNELS FROM SURVElLL"'l'fCE AllOW'AalE SPECIFIED REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.l 1.Main Sleam Une 15Clalil:m a.Read.<<Ve6&el Water 1.2.3 2'£I Btl 3.3.6.1.1 L3Q8 inches tete!..I.DN I.J:l'v Un'l.SiR 3.3.6.1.2 U!'l.ef 1 SR.:J...3.6.1.5 zero SR 3-:3.6.1.6 b.Main Sle;;m Lin:?reSEll"" 2 E SR:.3.6.1.2.IJ:AJc}SR:1.,3..6.1.5.siR 3.3.6.1.6 c.Main S1aam L\I1e Flow..1.2.3 2;:er 0 SR 33.fl:1.1$140%ra'.edMst.SR:1.,3..6.1.2&1eiimflal'l.siR:1.,3.6.1.5 SR 3.3.6.1.6 d.tv\1in Sleam T unoel 1.2,3 B[)SR:1.,3.6.1.2

$21JDof T efTlJefa!IJre

High SR 3.3.6.1:a SR:'.3.6.1.6 2.Plimar/

iEdalion a.ReactorVef>Sel Water 1.2.3 2 G SR 3.3.6.1.t inchel;wa-Lcw.I..ewl 3 SR:1.,3.6.1.2.siR 33.6.1.5 zero SR 3.3.6.1.6 b.Dlyi/e!l PrewJIl!-High 1.2.3 2 Btl:1.,3.6.1.2 i25psig SR 33.6.1.5 SR 3.3.6.1.6 3.High Prassur..Coooanl Injetiion (HPCI)S'yJ;tern l5Qlation a.HPOLire fIoI,,,*1.2.3 F SR 3;,3.6..1.2::;OOll6 i High SR:1<3.6.1.5 SiR 3.3.6.1.6 b.Hi'O 1.2.3 3 F SR 3..3.6.1.2 Pn!J;J;Ure

..Law SR 3..3.6.1.5 SR 3.3.6.1.6 c.HPCITt.mne 1.2.3 3 F SR 3.3.6.12:::20pslg EmallSt D':apt-Iagm SR 1oM.l.a?rewJre..High SR 3.3.6.1.6 3.3-59 Amendment No.263, 260, 296 September 14, 2006 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 fiXlge 2 of3}Primary Ccnlainmsnt 15Olalion In5lrumE!1rtalbn FUNcnC*N Af"?UCABLE MOOESOR OTHER SPECifiED CONDITIONS REQURED CH/f.NNELS PER TRIP SYSTEM CONDITIONS REfERENCED FROM REQUIRED ACTIONC-1

.:.tLOWASLE REQ.lnREMarrS VALUE 3.Hro Syv.em 15Olalion (ronlinued) d.HPCI Steam liM: Space Hl'C1 Pt.lll1p Rxlom Area

High e...Pel Ste-;;m UneSpace Torus (Exit)

-High f.HPa Steam lineSpace T(lfUl;Area Ter11JE!rarore

High g.HPel Ste-iim UI"F-Space Toruli I-rea (Emry)T en"flS!'aI.Iire

High 4.ReactorIscla1ioo 1.2.3 2 F SR 3.3.6.12 52n0"F1-3.6.1.5 SR.1-3.6.1.6 1.2;3 2 F SR 5180"1'SR 1,3.6.1.5 SR U.6.1.6 1.2.3 2 F SR 3.3.6.1.2$1S0"F SR 1,3.6.1.5 SR 3.:3.6.1.6 1,2.3 2 F SR 1,3.6.1.2 5180"F SR 1,3..6.1.5 SR 3.:3.6.1.6 a.RCIC&:am une AO'il-1.2.3 Hq1 F SR 54&J"H,O SR 1.3.6.1.5 SR 3,:3.6.1.6 F SR 3.3.6.1.2?fOp6lg SR 1,3.11.1.5 SR 3,.:3.6.1.6 F SR 1.3.6.1.2 SR.3..3.6.1.5 SR 3..3.6.1.6 F SR 3.3.6.1.2::;1811"F SR 3..3.6.1.5 SR 3.:3.6.1.6 F SR.1-3.6.12:>180"F SR 2...3.6.1.5 SR F SR 1.-3..6:12 SR 3.:3.6.1.5 SR F SR:1;;3.6.1.218G"F SR 3.3.6.1.5 SR 3.3.6.1.6 (CCfllinued}

2 2 2 1,2,3 1.2.3 1.2.3 1..2.3 e.RCIC Stem line Space T Ef'lllE!Iato'.lfe

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T er11Jerat>.re

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High d.RCIC&earn line Space RCIC Pt.ln1f>Rocm T er11Jefat\lre

-High c.ROle TUTbine Exhal.l5t iAaphragm?rewJre-High b.R'CIC&:am SoJppty Lir.e?r<!saJr!!

-Lew BFN-UNIT2 3.3-60 Amendment No. September 21, 2006

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE REVISION 43 PREPARED BY: RANDY WALDREP PHONE: 2038 RESPONSIBLE ORGANIZATION:

EMERGENCY PREPAREDNESS APPROVED BY: TONY ELMS EFFECTIVE DATE: 07/01/2008 LEVEL OF USE: REFERENCE USE QUALITY-RELATED DATE: 06/25/2008 REV.NO.REVISED PAGES HISTORY OF REVISION/REVIEW REASON FOR CURRENT REVISION (42 17-19 IC-53 BFN EPIP-1 revision 42 adjusts the information that supports EAL 1.1-G2 , 1.2-21,27,33,35, G and 1.5-S for changes resulting from engineering calculations that support 94,99,105, Minimum RPV Flooding Pressures (MRFP), and Heat Capacity Temperature 114,115, Limits.Revisions to these calculations were conducted for EOI Program 117,118, Manual Revision 27 (U2C15).The revision to the EOI Program manual adjusts 120,128,132 the EAL supportive information that is in compliance with the REP.EALs 2.3-A,2.3-S1,2.3-S2,2.3-G1,2.3-G2, 3.1-G, and 3.2-G were revised to adjust Unit 1 drywell radiation values to support the decision to not start Unit 1 at extended power up-rate (EPU).Calculations ND-N0090-930050 R11 and ND-N0090-930055 R12 support the conditions described above.The two calculations utilized to support the drywell radiation values are notafunction of the Emergency Operating Instruction.

This revision does not affect, alter or change the basis supporting the BFN TVA's standard emergency classification and action level scheme.Although this change does modify data and information utilized by existing EALs, the criteria established by NUREG 1.101 Rev.3 (NUMARC/NESP 007 Revision 2)concerning the development of emergency action levels are not modified or changed.Specific EALs utilize information/data maintained through the implementation/maintenance of the Emergency Operating Instruction (EO I)procedures as well as specific calculations thus establishing thresholds used as entry conditions for emergency classifications.

As calculations are revised based upon reactor parameters such as in this case, fuel specifications, the EAL threhold information must also be revised.This revision neither increases nor decreases the effectiveness of the REP.This revision simply adjusts data necessary to maintain the accuracy of applicable Emergency Action Levels.43 21,97,34, IC-54 Some pages were added which were intentionally left blank (and noted as so)75,127, to accommodate appropriate double sided printing and filing in procedure 129,188, manuals.189 EAL 1.2-A-Wording of EAL enhanced to clarify intent of EAL.EAL 7.3-U-Wording revised to change"greater than" to"exceeds or is predicted to exceed".Additionally, the basis page for this EAL addressed the escalation to the Alert classification.

This wording in the basis was also revised to change"greater than" to"exceeding or predicted to exceed".EAL 7.3-A-Wording in first condition changed from"greater than" to"exceeds or is predicted to exceed".Wording in second bullet of second condition changed from"Affecting equipment required for safe shutdown" to"Equipment required for safe shutdown is affected." Table 3.1-Removed from Table 3.1 the maximum safe operating temperature limit value for Core Spray BID Pump Room High Humidity or Temp High specific for Unit 2 and Unit 3.

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION TABLE OF CONTENTS EPIP*1 (TABLE OF CONTENTS 1 SECTION I

1.0 INTRODUCTION

3 1.1 Purpose 3

2.0 REFERENCES

3 2.1 Industry Documents 3 2.2 Plant Instructions 3 3.0 INSTRUCTIONS 4 3.1 Instructions 4 4.0 GLOSSARY of ABBREVIATIONS, ACRONYMS, AND DEFINITIONS 7 5.0 EVENT CLASSIFICATION INDEX 15 SECTION II EVENT CLASSIFICATION MATRiX 17 1.0 Reactor 17 2.0 Primary Containment 25 3.0 Secondary Containment 33 4.0 Radioactivity Release 39 5.0 Loss of Power 45 6.0 Hazards*51 7.0 Natural Events 69 8.0 Emergency Director Judgment 77 SECTION III BASiS 87 1.0 Reactor 87 2.0 Primary Containment 107 3.0 Secondary Containment 125 4.0 Radioactivity Release 134 5.0 Loss of Power 145 6.0 Hazards 155 7.0 Natural Events 180 8.0 Emergency Director Judgment 187 PAGE 1 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 THIS PAGE INTENTIONALLY BLANK PAGE 2 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPlp*1

1.0 INTRODUCTION

1.1 Purpose Provide guidance to the Shift Manager or Site Emergency Director (SED)for proper declaration and classification of emergencies and ensure emergency classifications are consistent with those used by state and local governments and the Nuclear Regulatory Commission (NRC).The procedure applies to site events that constitute an emergency consistent with those site specific events outlined in NUMARC/NESP-007 August 1992.The Shift Manager and the SED are the only persons authorized to make the emergency classification determination.

2.0 REFERENCES

2.1 Industry Documents A.NUREG-0654,"Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" B.10 CFR 50.47, Code of Federal Regulations C.Reg Guide 1.101 Rev.3,"Methodology for Development of Emergency Action Levels 2.2 Plant Instructions A.TVA Radiological Emergency Plan B.EPIP-2,"Notification of Unusual Event" C.EPIP-3,"Alert" D.EPIP-4,"Site Area Emergency" E.EPIP-5,"General Emergency" F.EPIP-16,"Termination and Recovery Procedure" PAGE 3 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 3.0 INSTRUCTION 3.1 Following plant events or transients review EPIP-1 Section II, 1.0 through 8.0 and determine if an event should be classified as an emergency.

NOTE 1.If an emergency action level for a higher classification was exceeded, but the present situation indicates a lower classification, the fact that the higher classification occurred shall be reported to the NRC and the CECC, if staffed, or ODS if the CECC is not staffed.The higher classification should not be declared.2.If an emergency action level was met but the emergency has been totally resolved, the emergency class that was appropriate shall be reported to the ODS and the NRC but should not be declared.3.1.1 EPIP-1 Section II, 1.0 through 8.0 captures events in eight major categories as listed on the event classification index.3.1.2 Each emergency action level (EAL)in a category is given annumeric designator.

The first numeric component of the EAL indicates the section followed by a numeric designator for the specific EAL within the section and an alpha numeric designator for the event class.Example: 5.2-U These designators provide for cross-reference between the specific EAL and the basis document which provides technical supporting information for the EAL and may aid the Shift Manager/SED in classifying events.Curves, notes, or tables that support the EAL are located on the face adjacent page within the matrix section of the procedure and are identified within the event classification window on the information bar that precedes the designator.

The information bar contains the appropriate indication to alert the user that a corresponding curve, note, or table applies to the EAL.Curves, notes, or tables that contain unit specific information will also be identified within the event classification windowbythe letter"US" located at the end of the EAL information bar.This information should alert the user that the corresponding curve, note, or table contains unit specific information.

Example I 5.2-U l CURVE I NOTE I TABLE 00 PAGE 4 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPlp*1 3.2 If the event is determined to be one of the four emergency classifications, the Shift Manager assumes the responsibility of SED until relieved by the Plant Manager or designee.3.2.1 Implement one ofthefollowing procedures as applicable:

EPIP-2 EPIP-3 EPIP-4 EPIP-5 Notification of Unusual Event Alert Site Area Emergency General Emergency 3.2.2 Continue to review the emergency conditions in the event classification matrix and escalate, terminate, or implement recovery as appropriate.

Refer to EPIP-16 for termination or recovery.3.3 If the event is determined not to be one of the four event classifications, continue to monitor plant conditions for possible changes that could result in reaching an event classification.

LAST TEXT PAGE 5 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 THIS PAGE INTENTIONALLY BLANK PAGE 6 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 4.0 GLOSSARY of ABBREVIATIONS, ACRONYMS, AND DEFINITIONS The following is a list of terms and phrases found in EPIP-1.Each term or phrase is provided with a meaning, to ensure consistent use and understanding.

TERM/PHRASE MEANING/DEFINITION ADS AOI Alert ARI ARM ARP ATWS Auto Bomb BWR Can/Cannot be determined Can/Cannot be Maintained Above/Below Can/Cannot be Restored Above/Below Automatic Depressurization System Abnormal Operating Instruction Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involve probably life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.Alternate Rod Insertion Area Radiation Monitor Alarm Response Procedure Anticipated Transient Without Scram Automatic An explosive device Boiling Water Reactor The current value or status of an identified parameter relative to that specified in the instruction can/cannot be ascertained using all available indications (direct and indirect, singly or in combination).

The value of the identified parameter(s) is/is not able to be kept above/below specified limits.This definition includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s)."Cannot" does not imply that the actual value of the parameter must first exceed the specified limit.The value of the identified parameter(s) is/is not able to be returned to above/below specified limits within a reasonable time after having exceeded the specified limits.This determination includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s).

PAGE 7 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION CAD CAS CDE CECC Ci Civil Disturbance Confinement Boundary CS deg DG Drywell EAl ECCS ECl EOI EPA EPIP EO Containment Atmosphere Dilution Central Alarm Station Committed Dose Equivalent Central Emergency Control Center Curie A group of 20 or more persons violently protesting station operations or activities at the site.Cubic Centimeters Spent Fuel Storage Cask CONFINEMENT BOUNDARY consisting of the MPC shell, bottom base plate, MPC lid (including the vent and drain port cover plates), MPC closure ring, and associated welds.Core Spray Degrees Diesel Generator The upper portion of the Primary Containment which encloses the Reactor Pressure Vessel.Emergency Action level Emergency Core Cooling System Effluent Concentration Limit Emergency Operating Instruction Environmental Protection Agency Emergency Plan Implementing Procedure Environmental Qualification PAGE 8 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP*1 TERM/PHRASE MEANING/DEFINITION Event Explosion F Fire Flammable Gas GOI General Emergency gm HCTL Hostage Hostile Action Assessment of an EVENT commences when recognition is made that one or more of the conditions associated with the event exists.Implicit in this definition is the need for timely assessment, i.e.within 15 minutes.A rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures required for safe operation.

Fahrenheit Combustion characterized by heat and light.Sources of smoke such as slipping drive belts or overheated electrical components do not constitute fires.Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.Combustible gasses maintained at concentrations less than the lower explosive limit.Will not explode due to ignition.General Operating Instruction Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels oftsite for more than the immediate site area.Gram Heat Capacity Temperature Limit A person(s)held as leverage against the station to ensure that demands will be met by the station.An act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end.This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force.Other acts that satisfy the overall intent may be included.Hostile Action should NOT be construed to be acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant.Non-terrorism based EALs should be used to address such activities, (e.g.violent acts between individuals in the owner controlled area).PAGE 9 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

HPCI High Pressure Coolant Injection HR Hour IN ISFSI Inches Independent Spent Fuel Storage Installation KV Kilovolt Large framed aircraft A large aircraft with the potential for causing significant damage to the plant;may be referred to as an airliner.LCO Limiting Condition for Operation LOCA Loss Of Coolant Accident LPCI Low Pressure Coolant Injection MRFP Minimum RPV Flooding Pressure MCUTL Maximum Core Uncovery Time Limit MIN Minute NA NA Avg.Reactor Coolant Temperature (oF)Reactor Mode Switch Position Title Power Operation Startup Run Refuel(a)or Startup/Hot Standby 3 Hot Shutdown(a)

Shutdown>2124Cold Shutdown(a)

Shutdown212 5 Refueling(b)

Shutdown or Refuel NA 1 2 Mode Modes of Operation (a)All reactor vessel head closure bolts fully tensioned.(b)One or more reactor vessel head closure bolts less that fully tensioned.

MPC Multi-Purpose Canister (part of ISFSI)PAGE 10 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPlp*1 TERM/PHRASE MEANING/DEFINITION MPH Miles per Hour mrem Millirem MSIV Main Steam Isolation Valve MSL Main Steam Line MSRV Main Steam Relief Valve NESP National Environmental Studies Project Notification of Events are in process or have occurred which indicate a potential Unusual Event degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.NUMARC Nuclear Management and Resources Council OCA Owner Controlled Area ODS Operations Duty Specialist 01 Operating Instruction OSC Operations Support Center PA Protected Area PAR Protective Action Recommendation PCIS Primary ContainmentIsolationSystem Primary The drywell, the vent system, and the suppression chamber.Containment Primary System Primary systems comprise the pipes, valves and other equipment connected to the RPV such that a reduction in RPV pressure will affect a decrease in the flow of steam or water being discharged through an unisolable break in the system.PAGE 11 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPlp*1 REVISION 43 TERM/PHRASE MEANING/DEFINITION Projectile An object ejected, thrown, or launched towards a plant structure.

The source of a projectile may be offsite or onsite.Damage is sufficient to cause concern regarding the integrity of the affected structure or the operability or reliability of safety equipment contained therein.Protected Area All areas within the security protected area fence.PSIG Pounds Per Square Inch Gauge R Rad RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System REP Radiological Emergency Plan RHR Residual Heat Removal RPS Reactor Protection System RPV Reactor Pressure Vessel Sabotage Deliberate damage, misalignment, misoperation of plant equipment with the intent to render equipment inoperable.

SAMG Severe Accident Management Guideline SEC Second Secondary The spaces immediately adjacent to or surrounding, the primary Containment containment from which the Reactor Building Ventilation System and the Standby Gas Treatment System provides a filtered elevated release.SED Site Emergency Director SGTS Standby Gas Treatment System Significant An unplanned event involving one or more of the following:

Transient (1)Automatic turbine runback greater than 25%thermal reactor power or (2)Electrical load reduction greater than 25%full electrical load, or (3)Thermal power oscillations greater than 10%, or (4)Reactor scram, or (5)Valid ECCS initiation.

PAGE 12 OF 206 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPlp*1 TERM/PHRASE MEANING/DEFINITION SI Site Area Emergency Site Boundary Subcritical Suppression Pool Suppression Chamber TAF TEDE Torus Toxic Gas TSC Valid Visible Damage Surveillance Instruction Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts (1)toward site personnel or equipment that could lead to the likely failure thereof or, (2)that prevent effective access to equipment needed for the protection of the public.Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.That line beyond which the land or property is not owned, leased, or otherwise controlledbyTVA.Reactor power below the heating range and not trending upward.The water volume contained in the suppression chamber intended to condense steam from an MSRV actuation or a primary system break inside the drywell, and provide an ECCS system injection water source.The structure enclosing the suppression pool water and the atmosphere above it.Top of Active Fuel Total Effective Dose Equivalent The lower portion of the primary containment which encloses the suppression pool.Equivalent to the suppression chamber.A gas that is dangerous to life or limb by reason of inhalation or skin contact.Technical Support Center An indication, report, or condition is considered to be valid when it is conclusively verified by redundant indicators or actual observation by plant personnel.

Damage to equipment that is readily observable without measurements, testing, or analysis.Damage is sufficient enough to cause concern regarding the continued operability or reliability of affected safety structure, system, or component.

PAGE 13 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 TERM/PHRASE MEANING/DEFINITION Vital Area WRGERMS yr An area that contains equipment necessary for the safe operations and shutdown of the plant.Wide Range Gaseous Effluent Radiation Monitoring System Year PAGE 14 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPIP-1 5.0 EVENT CLASSIFICATION INDEX SECTION 1.0 SECTION 2.0 SECTION 3.0 SECTION 4.0 SECTION 5.0 SECTION 6.0 SECTION 7.0 SECTION 8.0 REACTOR PRIMARY CONTAINMENT SECONDARY CONTAINMENT RADIOACTIVITY RELEASES LOSS OF POWER HAZARDS NATURAL EVENTS EMERGENCY DIRECTOR JUDGMENT 1.1 WATER LEVEL 1.2 SCRAM FAILURE 1.3 REACTOR COOLANT ACTIVITY 1.4 MSUOFFGAS RADIATION 1.5 LOSS OF DECAY HEAT REMOVAL 2.1 PRIMARY CONTAINMENT PRESSURE 2.2 PRIMARY CONTAINMENT HYDROGEN 2.3 DRYWELL RADIATION 2.4 DRYWELL INTERNAL LEAKAGE 2.5 LOSS OF PRIMARY CONTAINMENT 3.1 SECONDARY CONTAINMENT TEMPERATURE 3.2SECONDARYCONTAINMENT RADIATION 4.1 GASEOUS EFFLUENT 4.2 MAIN STEAM LINE BREAK 4.3 LIQUID EFFLUENT 5.1 LOSS OF AC POWER 5.2 LOSS OF 250V DC POWER 6.1 RADIOLOGICAL 6.2 CONTROL ROOM EVACUATION 6.3 TURBINE FAILURE 6.4 FIRE/EXPLOSION 6.5 TOXIC GASES 6.6 FLAMMABLE GASES 6.7 SECURITY 6.8 VEHICLE CRASH 6.9 SPENT FUEL STORAGE 7.1 EARTHQUAKE 7.2 TORNADO/HIGH WINDS 7.3 FLOOD 8.1 TECHNICAL SPECIFICATIONS 8.2 LOSS OF COMMUNICATION 8.3 LOSS OF ASSESSMENT CAPABILITY 8.4 OTHER LAST PAGE PAGE 15 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE INTRODUCTION EPlp*1 (THIS PAGE INTENTIONALLY BLANK PAGE 16 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX REACTOR 1.0 PAGE 17 OF 206 EPlp*1 REVISION 43 (BROWNS FERRY NOTES EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 1.1-U1/1.1-A1 1.1-S1 1.1-G2 CURVES/TABLES:

Applicable when the Reactor Head is removed and the Reactor CaVity is flooded.Applicable in Mode 5 when the Reactor Head is installed.

The reactor will remain subcritical under all conditions without boron when:*Unit 1: All control rods are inserted to or beyond position 02.Unit 2: Any 19 control rods are inserted to position 02, with all other control rods fUlly inserted.Unit 3: Any 19 control rods are inserted to position 02, with all other control rods fUlly inserted.*All control rods except one are inserted to or beyond position 00.*Determined by Reactor Engineering.

TABLE 1.1-G2 MINIMUM ALTERNATE RPV FLOODING PRESS (MARFP)NUMBER OF OPEN MSRVs MARFP (PSIG)6 or More 190 5 230 4 290 PAGE 18 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX WATER LEVEL EPIP-1 Description 1.1-U1 I I NOTEII Uncontrolled water level decrease in Reactor Cavity with irradiated fuel assemblies expected to remain covered by water.OPERATING CONDITION:

Mode 5 1.1-A1II NOTEII Uncontrolled water level decrease in Reactor Cavity expected to result in irradiated fuel assemblies being uncovered.

OPERATING CONDITION:

Mode 5 1.1-S1II NOTEII Reactor water level can NOT be maintained above-162 inches.(TAF)OPERATING CONDITION:

ALL 1.1-G1IIII Reactor water level can NOT be restored and maintained above-180 inches.OPERATING CONDITION:

Mode 1 or 2 or 3 Description 1.1-U2IIII Uncontrolled water level decrease in Spent Fuel Pool with irradiated fuel assemblies expected to remain covered by water.OPERATING CONDITION ALL 1.1-A2IIII Uncontrolled water level decrease in Spent Fuel Storage Pool expected to result in irradiated fuel assemblies being uncovered.

OPERATING CONDITION:

ALL 1.1-52IIII Reactor water level can NOT be determined.

OPERATING CONDITION:

Mode 1 or 2 or 3 1.1-G2II NOTE I TABLE I US Reactor water level can NOT be determined AND Either of the following exists:*The reactor will remain subcritical without boron under all conditions, andLess than 4 MSRVs can be opened, orReactor pressure can NOT be restored and maintained above Suppression Chamber pressure by at least.:.UNIT1-90 psi.:.UNIT2-80 psi.:.UNIT3-70 psi*It has NOT been determined that the reactor will remain subcritical without boron under all conditions and unable to restore and maintain MARFP in Table 1.1-G2.OPERATING CONDITION:

Mode 1 or 2 or 3 c z c C/)cr-m<m z-t C/)=i m m 3: m::u G)m z (")-<G)m z mr-mmm z (")-<PAGE 19 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES 1.2 Subcritical is defined as reactor power below the heating range and not trending upward.CURVESIT ABLES: UNIT 1 CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 14 15 SUPPR PL LVL (Fl)*__F/'OOVE CURVE FOR EXISTING RX PRESS UNIT2 CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 260 250 240 E 230 lL220....I 210 lL l!: 200 lL190 II)180 170 160 150 11.5 12 141516 SUPPR PL LVL 1FT)*ACTION UQUIUD IF A80VE CURVE FOR EXISTING RX PRESS 260 250 240 Ii:" 230....0.220....0.'" 0.0.iil 210 200 190 180 170 150-"OW'_............

_-...._...

.........11.51213 14 151617 18 19 SUPPR PL LVL (Fl)*AC'llCIII_1F

/'OOVE CURVE FOR EXISTING RX PRESS UNIT3 CURVE 1.2-G HEAT CAPACITY TEMP LIMIT PAGE 20 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP*1 SCRAM FAILURE REACTOR COOLANT ACTIVITY Description DescriptionIIII 1.3-UIIII Reactor coolant activity exceeds 26 J!Ci/gm dose equivalent 1-131 (Technical Specification Limits)as determined by chemistry sample.OPERATING CONDITION ALL c: z c: (I)c:)It.--m<m z-I 1.2-AII NOTEII Failure of RPS automatic scram functions to bring the reactor subcritical AND Manual scram or ARI (automatic or manual)was successful.

OPERATING CONDITION:

Mode 1 or2 1.3-AIIII Reactor coolant activity exceeds 300 J!Ci/gm dose equivalent lodine-131 as determined by chemistry sample.OPERATING CONDITION:

Mode 1 or 2 or 3)Itm1.2-SII NOTEII Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical.

OPERATING CONDITION:

Mode 1 1.2-G I CURVEIII US Failure of automatic scram, manual scram, and ARI.Reactor power is above 3%AND Either of the following conditions exists:*Suppression Pool temp exceeds HCTL.Refer to Curve 1.2-G.*Reactor water level can NOT be restored and maintained at or above-180 inches.OPERATING CONDITION:

Mode 1 or 2 I I I I I I I I (I)=i m m S m:::u (i)m z (')-<(j)m z m.--m s: mm z (')-<PAGE 21 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPlp*1 NOTES CURVESIT ABLES: UNIT 1 CURVE 1.5-8 HEAT CAPACITY TEMP LIMIT 14 15 SUPPR PL LVL (FT)*MmCIII__F I'BOVE CURVE FOR EXISTING RX PRESS 260 250 t 240 lI.230:IE 220 I!!..J 210 lI.0:: 200 lI.lI.190;:)til 180 170 160 W"l$150<11.5 12 13 14151617 SUPPR PL LVL 1FT)*ACTIOIIIU!QUIIlI!DIF ABOVE CURVE FOR EXISTING RX PRESS1819 19 18 17 16 13 260 250 240 G:'230....n.220 ill210 n.200 s:190 180 170 160 150 11.5 12 14 15 SUPPR PL LVL (FT)*MmCIII__lF ABOVE CURV.E FOR EXISTING RX PRESS PAGE 22 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 MSL/OFFGAS LOSS OF DECAY HEAT RADIATION REMOVAL Description Description 1.4-U IIIIIIII Valid MAIN STEAM LINE RADIATION HIGH-HIGH c: alarm, RA-90-135C Z c: OR (J)c:Valid OG PRETREATMENT RADIATION HIGH I alarm, RA-90-157A.

m<m OPERATING CONDITION:

Z Mode 1 or 2 or 3-I IIII 1.5-A IIII Reactor moderator temperature can NOT be maintained below 212 0 F whenever Technical Specifications require Mode 4 conditions or duringoperations in Mode 5.I m;;U-I OPERATING CONDITION:

Mode 4 or 5IIII 1.5-S I CURVE I I I US Suppression Pool temperature, level and RPV (J)pressure can NOT be maintained in the safe area=i of Curve 1.5-S.m m s: m;;U (i)m OPERATING CONDITION:

Z 0 Mode 1 or 2 or 3-<IIIIIIII (j)m z mI m s: mm z 0-<PAGE 23 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX THIS PAGE INTENTIONALLY BLANK PAGE 24 OF 206 EPIP-1 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 PRIMARY CONTAINMENT 2.0 PAGE 25 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES CURVESIT ABLES: TABLE 2.1-A INDICATIONS OF PRIMARY SYSTEM LEAKAGE INTO PRIMARY CONTAINMENT Primary Containment Pressure High Alarm Drywell Floor Drain Sump Pump Excessive Operation Drywell CAMActivityIncreasing Drywell Temperature Hiqh Alarm Chemistry Sample Radionuclide Comparison To Reactor Water 18 19 20 141516 17 SUPPR PL LVL (FT)SAFE CURVE 2.1-8 PRESS SUPPR PRESS O-F""'+-+---+----+-

.......-+--+--+--+--+-+-+-+-

1111.512 13 35 PAGE 26 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT PRESSURE HYDROGEN Description DescriptionIIIIIIII c: z c: C/)c:I""" m<m z-I 2.1-A I I I TABLE IIII I Drywell pressure at or above 2.45 psig ANDI""" Indication of Primary System leakage into m Primary Containment.

Refer to Table 2.1-A.::u-I OPERATING CONDITION:

Mode 1 or 2 or 3 2.1-S I CURVE III 2.2-S IIII Suppression Chamber pressure can NOT be Drywell or Suppression Chamber C/)maintained in the safe area of Curve 2.1-S.hydrogen concentration at or above 4%=i m AND m s: Drywell or Suppression Chamber m::u oxygen concentration at or above 5%.G)m OPERATING CONDITION:

OPERATING CONDITION:

Z C')Mode 1 or 2 or 3 Mode 1 or 2 or 3-<2.1-G IIII 2.2-G IIII Suppression Chamber pressure can NOT be Drywell or Suppression Chamber maintained below 55 psig.hydrogen concentration at or above 6%(i)m AND Z m Drywell or Suppression ChamberI""" oxygen concentration at or above 5%.m s: mm OPERATING CONDITION:

OPERATING CONDITION:

Z Mode 1 or 2 or 3 Mode 1 or 2 or 3 C')-<PAGE 27 OF 206 REVISION 43 BROWNS FERRY NOTES CURVESITABLES:

EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 TABLE 2.3-A/2.3-S2 DRYWELL RADIATION LEVELS WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT 3 RADMONITOR RlHR RADMONITOR RlHR RADMONITOR RlHR 1-RE-90-272A 196 2-RE-90-272A 642 3-RE-90-272A 196 1-RE-90-273A 297 2-RE-90-273A 297 3-RE-90-273A 297 TABLE 2.3-S1/2.3-G2 DRYWELL RADIATION LEVELS WITH RCS BARRIER NOT INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 UNIT 2 UNIT 3 RADMONITOR RlHR RADMONITOR RlHR RADMONITOR RlHR 1-RE-90-272A 2981 2-RE-90-272A 2263 3-RE-90-272A 2981 1-RE-90-273A 2960 2-RE-90-273A 2960 3-RE-90-273A 2960 TABLE 2.3-G1 DRYWELL RADIATION LEVELS WITH RCS BARRIER NOT INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 UNIT 2 UNIT 3 RADMONITOR RlHR RADMONITOR RlHR RADMONITOR RlHR 1-RE-90-272A 90091 2-RE-90-272A 68405 3-RE-90-272A 90091 1-RE-90-273A 89450 2-RE-90-273A 89450 3-RE-90-273A 89450 TABLE 2.3/2.S-U INDICATIONS OF LOSS OF PRIMARY CONTAINMENT Unexplained Loss Of Containment Pressure Exceedinq SI-4.7.A.2.a Limits Inabilitv To Isolate Anv Line Exitinq Containment When Isolation Is Required Ventinq Irrespective Of Offsite Release Rates Per EOls/SAMGs PAGE 28 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX DRYWELL RADIATION EPIP-1 2.3-AII I TABLE I US Drywell radiation levels at or above the values listed in Table 2.3-A/2.3-S2, with the RCS barrier intact inside Primary Containment.

Description IIII DescriptionIIII c: z c: CJ)c:>>r-m<m z-IIIII>>r-mOPERATING CONDITION:

Mode 1 or 2 or 3 2.3-S1III TABLE I US Drywell radiation levels at or above the values listed in Table 2.3-S1/2.3-G2 with the RCS barrier NOT intact inside Primary Containment.

OPERATING CONDITION:

Mode 1 or 2 or 3 2.3-G1III TABLE I US Drywell radiation levels at or above the values listed in Table 2.3-G1 with the RCS barrier NOT intact inside Primary Containment.

OPERATING CONDITION:

Mode 1 or 2 or 3 2.3-S2 I I I TABLE I US Drywell radiation levels at or above the values listed in Table 2.3-A/2.3-S2, with the RCS barrier intact inside Primary Containment, AND Either of the following exists:*Indications of loss of Primary Containment.

Refer to Table 2.3/2.5-U.

Primary Containment integrity can NOT be maintained.

OPERATING CONDITION:

Mode 1 or 2 or 3 2.3-G2III TABLE I US Drywell radiation levels at or above the values listed in Table 2.3-S1/2.3-G2 with the RCS barrier NOT intact inside Primary Containment, AND Either of the following exists:*Indications of loss of Primary Containment.

Refer to Table 2.3/2.5-U.

Primary Containment integrity can NOT be maintained.

OPERATING CONDITION:

Mode 1 or 2 or 3 CJ)=i m m Ri::u Ci)m z o-<G)m z mr-m s: mm z o-<PAGE 29 OF 206 REVISION 43 BROWNS FERRY NOTES CURVESITABLES:

EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP*1 TABLE 2.3/2.5-U INDICATIONS OF LOSS OF PRIMARY CONTAINMENT Unexplained Loss Of Containment Pressure ExceedinQ SI-4.7.A.2.a Limits Inabilitv To Isolate Anv Line ExitinQ Containment When Isolation Is Required VentinQ Irrespective Of Offsite Release Rates Per EOls/SAMGs PAGE 30 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 DRYWELL INTERNAL LOSS OF PRIMARY LEAKAGE CONTAINMENT Description Description 2.4-U IIII 2.S-U I I I TABLE I Drywell unidentified leakage exceeds 10 gpm Inability to maintain Primary Containment c: pressure boundary.Refer to Table 2.3/2.5-U.

Z OR c: (t)Drywell identified leakage exceeds 40 gpm.c:>>r-m OPERATING CONDITION:

OPERATING CONDITION:

<: m Mode 1 or 2 or 3 Mode 1 or 2 or 3 Z-I 2.4-A IIIIIIII Drywell unidentified leakage exceeds 50 gpm.>>r-mOPERATING CONDITION:

Mode 1 or 2 or 3IIIIIIII (t)=i m m S m::u Ci)m z 0-<IIIIIIII (j)m z mr-m s:: mm z (")-<PAGE 31 OF 206 REVISION 43 (BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX THIS PAGE INTENTIONALLY BLANK PAGE 32 OF 206 EPIP-1 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 SECONDARY CONTAINMENT 3.0 PAGE 33 OF 206 REVISION 43 BROWNS FERRY NOTES CURVESITABLES:

EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPlp*1 (TABLE 3.1 MAXIMUM SAFE OPERATING AREA TEMPERATURE LIMITS APPLICABLE PANEL 9-21 MAX SAFE OPERATING AREA TEMPERATURE ELEMENTS VALUE of (UNLESS OTHERWISE NOTED)UNIT 1 UNIT 2 UNIT3 RHR AlC Pump Room 74-95A 215 150 155 RHR BID Pump Room 74-95B 150 210 215 HPCI Turbine Area 73-55A 275 270 270 CS AlC Pump and RCIC Turbine Area 71-41A 190 190 190 RCIC Steam Supply Area 71-41B, 41C, 410 195 200 250 HPCI Steam Supply Area 73-55B, 55C, 55D 245 240 240 RHR AlC Pump Supply Area 74-95H 245 240 240 RHR BID Pump Supply Area 74-95G 190 240 240 Main Steam Line Leak Detection High (XA-55-3D-24)

Panel 9-3 TIS-1-60A 315315315 RHR Valve Room 74-95E 175 170 175 RWCU 1501 Logic Channel AlB Temp (XA-55-5B-32/33)

Panel 9-5 175 170 175 High 69-835A, B, C, D Aux Inst Room RWCU Outbd 1501 Vlv Area 69-29F 220 220 220 RWCU HxArea 69-29G 220 220 220 RWCU Hx Exh Duct 69-29H 220220220 RWCU Recirc Pump A Area 69-29D 215 215 215 RWCU Recirc Pump B Area 69-29E 215 215 215 RHR AlC Hx Room 74-95C 210 195 200 RHR BID Hx Room 74-95D 210 195 200 FPC HxArea 74-95F160155 155 TABLE 3.1-G/3.2-G INDICATIONS OF POTENTIAL OR SIGNIFICANT FUEL CLADDING FAILURE WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 DRYWELL RADIATION 1-RE-90-272AI>196 RlHR 2-RE-90-272A I>642 RlHR 3-RE-90-272A I>196 RlHR 1-RE-90-273AI>297 RlHR 2-RE-90-273A I>297 RlHR 3-RE-90-273A I>297 RlHR Reactor Coolant Activity Reactor Coolant Activity Reactor Coolant Activity300 IlCi/gm Dose Equivalent300 IlCi/gm Dose Equivalent300 IlCi/gm Dose Equivalent Iodine 131 Iodine 131 Iodine 131 PAGE 34 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 SECONDARY CONTAINMENT TEMPERATURE DescriptionIIIII c: z c: (t)c:>>r-m<m z-IIIIII>>r-m3.1-S I I I TABLE I US I An unisolable Primary System leak is discharging into Secondary Containment AND Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1.OPERATING CONDITION:

Mode 1 or 2 or 3 (t)=i m m s: m;;0 Ci)m z o-<3.1-G I I I TABLE I US I An unisolable Primary System leak is discharging into Secondary Containment AND Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1 AND Any indication of potential or significant fuel cladding failure exists.Refer to Table 3.1-G/3.2-G with RCS Barrier intact inside Primary Containment.

OPERATING CONDITION Mode 1 or 2 or 3 (j)m z mr-mmm z o-<PAGE 35 OF 206 REVISION 43 (BROWNS FERRY NOTES CURVESIT ABLES: EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 TABLE 3.2 MAXIMUM SAFE OPERATING AREA RADIATION LIMITS AREA RADMONITOR MAX SAFE VALUE MRlHR RHR West Room 90-25A 1000 RHR East Room 90-28A 1000 HPCIRoom 90-24A 1000 CS/RCIC Room 90-26A 1000 Core Spray Room 90-27A 1000 Suppr Pool Area 90-29A 1000 CRD-HCU West Area 90-20A 1000 CRD-HCU East Area 90-21A 1000 TIP Drive Area 90-23A 1000 North RWCU System Area 90-13A 1000 South RWCU System Area 90-14A 1000 RWCU System Area 90-9A 1000 MG Set Area 90-4A 1000 Fuel Pool Area 90-1A 1000 Service Fir Area 90-2A 1000 New Fuel Storaqe 90-3A 1000 TABLE 3.1-G/3.2-G INDICATIONS OF POTENTIAL OR SIGNIFICANT FUEL CLADDING FAILURE WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 DRYWELL RADIATION 1-RE-90-272AI>196 RlHR 2-RE-90-272A I>642 RlHR 3-RE-90-272A I>196 RlHR 1-RE-90-273AI>297 RlHR 2-RE-90-273A I>297 R/HR 3-RE-90-273A I>297 RlHR Reactor Coolant Activity Reactor Coolant Activity Reactor Coolant Activity 2: 300 J.lCilgm Dose Equivalent 2: 300 J.lCilgm Dose Equivalent 2: 300 J.lCilgm Dose Equivalent Iodine 131 Iodine 131 Iodine 131 PAGE 36 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 SECONDARY CONTAINMENT RADIATION III DescriptionII c: z c: CJ)c:....m<m z.....3.2-AIIIII Any of the following high radiation alarms on Panel 9-3:*RA-90-1A, Fuel Pool Floor Alarm*RA-90-250A, Reactor, Turbine, Refuel Exhaust*RA-90-142A, Reactor Refuel Exhaust*RA-90-140A, Refueling Zone Exhaust AND Confirmation by Refuel Floor personnel that irradiated fuel damage mayhaveoccurred.

OPERATING CONDITION:

ALL 3.2-SIII TABLEII An unisolable Primary System leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2.OPERATING CONDITION:

Mode 1 or 2 or 3 3.2-GIII TABLE I US I An unisolable Primary System leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2.AND Any indication of potential or significant fuel cladding failure exists.Refer to Table 3.1-G/3.2-G with RCS Barrier intact inside Primary Containment.

OPERATING CONDITION Mode 1 or 2 or 3 CJ)=i m m s: m;U G)m z o-<G)m z m...m s: mm z o-<PAGE 37 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX THIS PAGE INTENTIONALLY BLANK PAGE 38 OF 206 EPIP-1 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 RADIOACTIVITY RELEASES 4.0 PAGE 39 OF 206 REVISION 43 BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 NOTES 4.1-U Prior to making this emergency dassification based upon the WRGERMS indication, assess the release by either of the folloVving:

1.Actual field measurements exceed the limits in table 4.1-U 2.D-SI 4.8.B.1.a.1 release fraction exceeds 2.0 If neither assessment can be conducted Vvithin 60 minutes then the declaration must be made on the valid WRGERMS reading.4.1-A Prior to making this emergency dassification based upon the WRGERMS indication, assess the release by either of the folloVving:

1.Actual field measurements exceed the limits in table 4.1-A 2.D-S14.8.B.1.a.1 release fraction exceeds 200 If neither assessment can be conducted Vvithin 15 minutes then the declaration must be made on the valid WRGERMS reading.4.1-5 Prior to making this emergency dassification based upon the gaseous release rate indication, assess the release by either of the folloVving methods: 1.Actual field measurements exceed the limits in table 4.1-5.2.Projected or actual dose assessments exceed 100 mrem TEDE or 500 mrem CDE.If neither assessment can be conducted Vvithin 15 minutes then the dedaration must be made based on the valid WRGERMS reading.4.1-G Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the folloVving methods: 1.Actual field measurements exceed the limits in table 4.1-G.2.Projected Of actual dose assessments exceed 1000 mrem TEDE or 5000 mrem CDE.If neither assessment can be conducted Vvithin 15 minutes then the dedaration must be made based on the valid WRGERMS reading.CURVESITABLES' Table 4.1-U RELEASE LIMITS FOR UNUSUAL EVENT TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS)2.88 X 10 7!LCi/sec 1 Hour Gaseous Release Rate O-SI 4.8.B.1.a.1 Release Fraction 2.0 1 Hour Site Boundary Radiation Reading Field Assessment Team 0.10 MREM/HR Gamma 1 Hour Table 4.1-A RELEASE LIMITS FOR ALERT TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS)2.88 X 10 9!LCi/sec 15 Minutes Gaseous Release Rate 0-SI4.8.B.1.a.1 Release Fraction 200 15 Minutes Site Boundary Radiation Reading Field Assessment Team 10 MREM/HR Gamma 15 Minutes Table 4.1-S RELEASE LIMITS FOR SITE AREA EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS)5.9 X 10 9!LCilsec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 100 MREM/HR Gamma 1 Hour Site Boundary lodine-131 Field Assessment Team 3.9 X 10-7!LCI/cm 3 1 Hour Table 4.1-G RELEASE LIMITS FOR GENERAL EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS)5.9 X 10 10 15 Minutes Site Boundary Radiation Reading Field Assessment Team 1000 MREM/HR Gamma 1 Hour Site Boundary lodine-131 Field Assessment Team 3.9 X 10-6cm 3 1 Hour PAGE 40 OF 206 REVISION 43 BROWNS FERRY 4.1-U I EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX GASEOUS EFFLUENT Description I NOTE I TABLEII EPIP-1 Gaseous release exceeds ANY limit and duration in Table 4.1-U.OPERATING CONDITION:

ALL c: z c: (J)c:>r-m<m z-I 4.1-A I I NOTE I TABLE I I Gaseous release exceeds ANY limit and duration in Table 4.1-A.OPERATING CONDITION:

ALL 4.1-S I I NOTE I TABLEII EITHER of the following conditions exists:*Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-5.*Dose assessment indicates actual or projected dose consequences above 100 mrem TEDE or 500 mrem thyroid CDE.OPERATING CONDITION:

ALL 4.1-GII NOTE I TABLEII EITHER of the following conditions exists:*Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-G.*Dose assessment indicates actual or projected dose consequences above 1000 mrem TEDE or 5000 mrem thyroid CDE.OPERATING CONDITION ALL PAGE 41 OF 206>r-m(J)=i m mm z (')-<(j)m z mr-m s: mm z (')-<REVISION 43 (BROWNS FERRY NOTES CURVESIT ABLES: EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX PAGE 42 OF 206 EPIP-1 REVISION 43

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