Text

05202014 Final Browns Ferry 2014-301 SRO Written Exam
	ML14142A161
Person / Time
Site:	Browns Ferry
Issue date:	05/21/2014
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
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	Download: ML14142A161 (115)
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NRC FORM 665P U.S. NUCLEAR REGULATORY COMMISSION PACKAGE (Rev. 3-2009)

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FACILITY NAME: BROWNS FERRY Section 9 REPORT NUMBER: 05000259/260/296 2014-301 FINAL SRO WRITTEN EXAM CONTENTS:

/

D Final SRO Written Exam (25 as given questions with changes made during administration annotated)

O Reference Handouts Provided To Applicants El Answer Key Location of Electronic Files:

Submitted By: -

Lti Verified By:

NRC Exam Site-Specific SRO Written Examination Applicant Information Name:

Date: Facility/Unit: Browns Ferry Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

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QUESTION 76 Unit 2 is operating at 100% Reactor Power, when the following alarms are received:

HPCI LOGIC POWER FAILURE, (2-9-3 F, Window 3)

HPCI 120 VAC POWER FAILURE, (2-9-3F, Window 7)

ADS BLOWDOWN POWER FAILURE, (2-9-3C, Window 32)

CORE SPRAY SYS II LOGIC PWR FAILURE, (2-9-3 F, Window 23)

RHR SYS II LOGIC POWER FAILURE, (2-9-3E, Window 5)

Which ONE of the following completes both statements below?

The 250 VDC RMOV Board (1) has been lost.

In accordance with Tech Spec Bases 3.8.7, Distribution Systems Operating, if this RMOV Board is transferred to its alternate source, it (2).

A. (l)2A (2) must be declared inoperable B. (1)2A (2) remains OPERABLE C. (1)2B (2) must be declared inoperable D. (1)2B (2) remains OPERABLE

QUESTION 77 A reactor scram has occurred on Unit 1.

Current plant conditions are:

Mode Switch: SHUTDOWN

17 Control Rods are at position 02

RPV level: (+) 31 inches

Reactor Pressure: 955 psig

Drywell Pressure: 4 psig and rising

DRYWELL RADIATION HIGH, (1-9-7C, Window 15), is alarming AND verified valid

Standby Liquid Control is injecting Which ONE of the following completes both statements below?

Based on the current plant conditions, control rod insertion is directed in accordance with (1).

If the Unit Operator subsequently reports that ALL control rods are fully inserted, the Unit Supervisor will (2)_.

A. (1) 1-AOI-100-1, Reactor Scram (2) stop Boron injection B. (1) 1-AOl-lOU-i, Reactor Scram (2) continue Boron injection C. (1) 1-EOI-1, RPV CONTROL, RC/Q (2) stop Boron injection D. (1) 1-EOI-1, RPV CONTROL, RC/Q (2) continue Boron injection

QUESTION 78 Given the following conditions for Unit 2:

Reactor has scrammed due to a loss of control air header pressure.

Suppression pool temperature is 1150 F

Control Air header pressure is 25 psig

RPV water level was deliberately lowered to (-) 55 inches Which ONE of the following completes the statements below?

MSIV status is (1).

_(2)_ contains the steps to re-align Drywell Control Air to the MSIVs.

NOTE: 2-AOI-32-2, Loss of Control Air 2-EOI Appendix-8B, Reopening MSIVs/Bypass Valve Operation A. (1) all MSIVs CLOSED (2) 2-AOI-32-2 B. (1) all MSIVs CLOSED (2) 2-EOT Appendix-8B C. (1) inboard MSIVs OPEN, outboard MSIVs CLOSED (2) 2-AOI-32-2 D. (1) inboard MSIVs OPEN, outboard MSIVs CLOSED (2) 2-EOI Appendix-8B

QUESTION 79 Unit 1 was at 100% power when one Safety Relief Valve failed open and was unable to be closed. The Reactor Mode Switch was placed in SHUTDOWN.

The following conditions exist:

Reactor power 8% and lowering

Reactor pressure 900 psig and stable

Suppression pooi temperature 190 °F and rising

Suppression pooi level 16.0 ft and slowly rising Which ONE of the following identifies the required action in accordance with 1-EOI-2, Primary Containment Control?

REFERENCE PROVIDED A. lower reactor pressure but maintain cool down rate limitations, in accordance with 1-EOI-1, RPV Control B. lower reactor pressure irrespective of cool down rate, in accordance with 1-EOI-1, RPV Control C. anticipate Emergency Depressurization in accordance with 1 -C-2, Emergency RPV Depressurization D. Emergency Depressurize in accordance with 1-C.2, Emergency RPV Depressurization

QUESTION 80 Unit 1 was operating at 100% power, when an ATWS occurred with the following conditions:

Reactor Power is unknown

RPV pressure is being maintained between 800 to 1000 psig

Suppression Pool temperature is 102 °F and stabilizing

RHR Suppression Pool cooling is in service in accordance with 1-EOI Appendix-17A, RHR System Operation Suppression Pool Cooling Which ONE of the following completes both statements below?

In accordance with Emergency Operating Instructions, the RPV Level Band is _(1)_ inches.

In accordance with 1 -EOI- 1, RPV Control, implementation of Appendix-3A, SLC Injection _(2)_ required.

A. (1) (+) 51 to (-) 180 (2) is B. (1) (+) 51 to (-) 180 (2) is NOT C. (1)(-)5Oto(-) 180 (2) is D. (1)(-)5Oto(-)180 (2) is NOT

QUESTION 81 The following conditions exist for Unit 1:

Mode3

Reactor Pressure 75 psig

Core Spray Loop 2 is INOPERABLE

RHR Loop 2 is operating in Shutdown Cooling Which ONE of the following identifies a subsequent inoperability that, in combination with the conditions listed above, requires an hourly fire watch?

REFERENCE PROVIDED A. Preaction System for HPCI, 1-26-37 B. Unit 1 Auxiliary Instrument Room C02 System C. Fire Detection for RHR on Panel 1-LPNL-25-545 D. Fire Detection for RCIC on Panel 1-LPNL-25-545

QUESTION 82 All three units are at 100% power when the following occurs:

0-AOI-57-.1E,Grid Instability, has been entered

The Transmission Operator (TOp) reports the status of offsite power sources:

o 161 kV is U11QUALIFIED o 500 kV is QUALFIED

BOTH starting air receivers are at 167 psig for the 3A Diesel Generator Which ONE of the following completes the statement below?

The MOST limiting condition for Unit 3 in accordance with T.S. 3.8.1, AC Sources Operating, is condition REFERENCE PROVIDED A. A B. E C. F D. J

QUESTION 83 Given the following conditions:

A loss of coolant accident (LOCA) has occurred on Unit 1 Suppression Pool level is 16 feet Drywell Pressure is 25 psig

/0 concentrations are as indicated below:

H 2

Which ONE of the following completes the statements below?

In accordance with 1 -EOI-Appendix- 19, 2 /0 Analyzer Operation, readings from 1 -XR 110 H

/0 Concentration Recorder (Panel 1-9-54) or from 1 -MON 110, H H

2 /O Analyzer 2

(Panel 1-9-55) (1).

Based on the given parameters, in accordance with EPTP-1, Emergency Classification the current classification is a(an) (2).

REFERENCE PROVIDED A. (1) are valid as soon as the analyzer is placed in service (2) Alert B. (1) are valid as soon as the analyzer is placed in service (2) Site Area Emergency C. (1) are NOT valid until ten minutes after the analyzer is placed in service (2) Alert D. (1) are NOT valid until ten minutes after the analyzer is placed in service (2) Site Area Emergency

QUESTION 84 Unit 1 has experienced a LOCA and the following containment parameters exist:

Drywell Pressure is 23.4 psig

Suppression Chamber Pressure is 22 psig

Hydrogen concentration in the Drywell is 2.9%

Suppression Pool Level is 15 feet

Reactor Water Level is (-) 170 inches and rising Which ONE of the following completes the statements below?

The required procedure is _( 1 )_.

Vent under these conditions _(2)_.

A. (1) 1-EOIAPPENDIX-12, Primary Containment Venting (2) irrespective of offsite radioactive release rates B. (1) 1-EOIAPPENDIX-12, Primary Containment Venting (2) ONLY if offsite radioactive release rates can be maintained below ODCM limits C. (1) 1-EOIAPPENDIX-15, RPV Venting for Primary Containment Flooding (2) irrespective of offsite radioactive release rates D. (1) 1-EOIAPPENDIX-15, RPV Venting for Primary Containment Flooding (2) ONLY if offsite radioactive release rates can be maintained below ODCM limits

QUESTION 85 Unit 1 was operating at 100% Reactor Power when a Reactor Scram occurred.

Following the Scram a primary system leak into Secondary Containment, with the following alarms and indications:

RX BLDG AREA RADIATION HIGH, (1-9-3A, Window 22)

RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH, (1-9-3A, Window 4)

Elevation 565 East ARM meter indicating off-scale high

Elevation 565 Northeast ARM meter indicating 600 mr/hr and stable

An Unusual Event has been declared based on Stack Noble Gas (WRGERMS) exceeding 2.88 X 10 iiCi/sec for more than an hour.

Which ONE of the following completes the statement?

In accordance with 1-EOI-3, Secondary Containment Control, the crew is required to enter (1)

A potential isolation source in accordance with Table 4, Secondary Cntmt Area Radiation is (2).

(1) (2)

A. 1 -EOI- 1, RPV Control SDV vents and drains B. 1-EOI-1, RPV Control RWCU suction and return isolation valves C. 0-EOI-4, Radioactivity Release Control SDV vents and drains D. 0-EOI-4, Radioactivity Release Control RWCU suction and return isolation valves

QUESTION 86 Unit 2 was shutdown due to RHR Loop II being inoperable for 6 days.

RI-JR Loop II is tagged out for maintenance on the 2-FCV-74-66, Outboard LPCI Injection Valve

RHR Loop I is in Shutdown Cooling with RHR pump 2A Which ONE of the following completes both statements below?

In accordance with Technical Specification 3.5.2, ECCS Shutdown, for these plant conditions,

_(1)_ low pressure ECCS injection/spray subsystem(s) is(are) required to be Operable.

While aligned for Shutdown Cooling, RI-JR pump 2A _(2)_ be considered an Operable ECCS subsystem.

A. (1)one (2) can B. (l)one (2) CANNOT C. (1)two (2) can D. (1)two (2) CANNOT

QUESTION 87 Unit 3 has entered Mode 1 two hours ago.

During Instrument Maintenance (IM) testing on the APRM 2-Out-Of-4 Voters, Voter 1 failed and the IMs have determined that Voter 1 will NOT generate an output signal to RPS.

During troubleshooting activities Engineering and IMs have determined that due to a common mode failure 2-Out-Of-4 Voter 3 also will NOT generate an output signal to RPS.

Which ONE of the following identifies the MOST limiting Technical Specification required action?

REFERENCE PROVIDED A. Required Action A. 1 OR A.2 with a Completion Time of 12 Hours B. Required Action B. 1 OR B.2 with a Completion Time of 6 Hours C. Required Action C. 1 with a Completion Time of 1 Hour D. Required Action G. 1 with a Completion Time of 12 Hours

QUESTION 88 Given the following plant conditions on May 5, 2014 at 08:00:

Unit 1 is starting up, currently in Mode 2

Unit 2 is in Mode 1, 100% power

Unit 3 is in Mode 5 with core unloading in progress

Diesel Generator 3ED is inoperable At 09:00, SGT Train B is declared inoperable.

Which ONE of the following identifies whether further fuel movement is allowed and the EARLIEST time Unit 1 and 2 must be placed in Mode 3?

REFERENCE PROVIDED A. Fuel moves may continue; May 5 at 23:00 B. Fuel moves may continue; May 6 at 02:00 C. Fuel moves may NOT continue; May 5 at 23:00 D. Fuel moves may NOT continue; May 6 at 02:00

QUESTION 89 Unit 3 is at 100% power and the following condition exists:

Unit Battery 3 is on a float charge and ventilation for this area has been lost

Ambient room temperature at Unit 3 Battery is 103 °F

All other battery cell parameters are normal Which ONE of the following completes both statements below?

Unit 3 Battery is _(1)_.

The procedure which provides guidance for setting up temporary ventilation to the area where this battery is located is (2).

A. (1) OPERABLE (2) 0-01-3 OF, Common and DG Building Ventilation B. (1) OPERABLE (2) 0-01-31, Control Bay and Off-Gas Treatment Building Air Conditioning System C. (1) INOPERABLE (2) 0-01-3 OF, Common and DG Building Ventilation D. (1) INOPERABLE (2) 0-01-31, Control Bay and 0ff-Gas Treatment Building Air Conditioning System

QUESTION 90 Unit 1 is Mode 5 with vessel reassembly in progress with gates installed. RHR Pump 1A is operating in Shutdown Cooling.

RBCCW Pump lB has tripped

RBCCW PUMP SUCTION HDR TEMP has increased to 106° F

Spare RBCCW Pump is UNAVAILABLE RWCU System AND the Fuel Pool Cooling System have been shutdown as directed by 1 -AOI 1, Loss of Reactor Building Closed Cooling Water.

NOTE: Section 8.14, Initiation of Supplemental Fuel Pool Cooling with RHR Drain Pump B Section 8.16, Initiation of Supplemental Fuel Pool Cooling with RHR Pumps BorD Which ONE of the following completes the statements below?

The Fuel Pool temperature limit in TRM 3.9.2 is _(1)_.

In accordance with 1-01-74, RHR System, the preferred method of supplemental fuel pool cooling is _(2)_.

A. (1) 125° F (2) Section 8.14 B. (1) 125° F (2) Section 8.16 C. (1) 150° F (2) Section 8.14 D. (1) 150° F (2) Section 8.16

QUESTION 91 Unit 1 was operating at 75% power at 70% core flow, when the Reactor Recirc Loop B flow transmitter, 1 -FT-68-8 1 C, input to APRM 3 fails downscale.

Which ONE of the following completes both statements below?

A Control Rod Block (1) occur.

In accordance with Technical Specification 3.3.2.1, Control Rod Block Instrumentation, Action Condition A _(2)_.

REFERENCE PROVIDED A. (1)willNOT (2) is NOT required at this time B. (1)willNOT (2) is required at this time C. (1) will (2) is NOT required at this time D. (l)will (2) is required at this time

QUESTION 92 Unit 2 Mode Switch is in Startup/Hot Standby, Reactor Power is 8%, and preparations are being made to place the Mode Switch in Run.

Subsequently:

A failure in Main Turbine Bypass Valve control has caused Reactor Pressure to rise

Attempts to manually control Main Turbine Bypass Valves were unsuccessful

The Reactor was MANUALLY scrammed when pressure was 1060 psig and rising Which ONE of the following completes both statements below?

Technical Specification 3.4.10, Reactor Steam Dome Pressure, Limiting Condition for Operation (1) exceeded.

The FIRST Immediate NRC Notification required is a _(2)_ report in accordance with NPG-SPP-03 .5, Regulatory Reporting Requirements.

REFERENCE PROVIDED A. (1) was (2) 4-Hour B. (1) was (2) 8-Hour C. (1)wasNOT (2) 4-Hour D. (1)wasNOT (2) 8-Hour

QUESTION 93 Unit 2 is operating at 100% power with hydrogen water chemistry in service.

Subsequently:

2-FCV-66-14 and 18, Off-Gas Outlet Valve from SJAE A and B automatically close.

Which ONE of the following completes both statements below?

The reason why hydrogen concentration is expected to initially rise is because _(1)_.

The required action if hydrogen concentration reaches and remains above 4.1% in accordance with 2-AOI 1, Off-Gas H2 High is to perform _(2)_.

A. (1) HWC is still in service (2) 2-GOI-100-12, Power Maneuvering B. (1) HWC is still in service (2) 2-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions In Power During Power Operations C. (1) HWC has isolated (2) 2-GOT- 100-12, Power Maneuvering D. (1) HWC has isolated (2) 2-GOT- 100-1 2A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions In Power During Power Operations

QUESTION 94 Unit 1 is at 8% power performing 1-GOI-100-1A, Unit Startup, and preparing to transition to MODE 1.

The following alarm and pump seal pressure indications are received:

RECIRC PUMP 1A NO 1 SEAL LEAKAGE ABN (1-9-4A, Window 25)

No. 1 Seal Pressure: 980 psig

No. 2 Seal Pressure: 980 psig In accordance with Tech Specs, which ONE of the following completes the statements below?

The alarm/indications (1) RCS pressure boundary leakage.

Mode 1 (2) be entered if the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period average for TOTAL operational leakage is 31 gpm.

A. (1) signify (2) can B. (1) signify (2) CANNOT C. (1) do NOT signify (2) can D. (1) do NOT signify (2) CANNOT

QUESTION 95 ACTIONS CONDfl1ON REQUIRED ACTION COMPLETION TIME A One pump AJ Restore pump 7 days inoperable, to OPERABLE status.

B. Required a,1 Be in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action and MODE 3.

associated Completion ANP Time not met B.2 Bern 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months MODE 4.

Given the above LCO:

th 1

Pump 1 becomes inoperable at 0600 on May 1 th*

12 Pump 2 becomes inoperable at 0100 on May Which ONE of the following completes the statements below?

(Both pumps are in the same system)

If Pump 1 is restored to OPERABLE at 0800 on May 12, and if Pump 2 is NOT restored to OPERABLE then Condition B is entered at time (1).

th, 12 If Pump 2 is restored to OPERABLE at 0800 on May and if Pump 1 is NOT restored to OPERABLE then Condition B is entered at time _(2)_.

th 19 A. (1) 0100 May th 18 (2) 0600 May B. t(1)0100May19 (2) 0600 May 19 th th 20 C. (1) 0100 May th 18 (2) 0600 May th 20 D. (1) 0100 May th 19 (2) 0600 May

QUESTION 96 Unit 3 was operating at 100% Reactor Power. RHR Pump 3B was tagged out for planned maintenance at 0600 on 1/13/14.

At 1000 on 1/14/14, a trip of Loop I Core Spray Room Cooler occurred.

Based on these conditions, which ONE of the following identifies the LATEST time that Unit 3 must be in Mode 3 in accordance with Tech Spec 3.5.1, ECCS-Operating?

REFERENCE PROVIDED A. 2200on 1/14/14 B. 2300 on 1/14/14 C. l800on 1/20/14 D. 2200on1/21/14

QUESTION 97 Which ONE of the following completes both statements below in accordance with O-SI-4.8.A.1-1, Liquid Effluent Permit?

A release that is to be performed with an inoperable O-RM-90-130, Radwaste Effluent Monitor, is required to be authorized by the _(l)_.

If the O-RM-90-130 monitor is declared inoperable during a release, then the release (2).

A. (1) Chemistry Manager and the Unit Supervisor (2) may continue if an independent verification of the valve lineup and release calculation is immediately performed B. (1) Chemistry Manager and the Unit Supervisor (2) must be terminated and a new release permit initiated C. (1) Unit Supervisor ONLY (2) may continue if an independent verification of the valve lineup and release calculation is immediately performed D. (1) Unit Supervisor ONLY (2) must be terminated and a new release permit initiated

QUESTION 98 Which ONE of the following completes the statements below?

In accordance with the Technical Specification Bases of 3.9.1, Refueling Equipment Interlocks, with one or more required refueling equipment interlocks inoperable, in-vessel fuel movements

_( 1 )_.

In accordance with O-GOI-100-3C, Fuel Movement Operations During Refueling, work above Drywell elevation _(2)_ during fuel movements is controlled in accordance with RCI- 17, Control of High Radiation Areas and Very High Radiation Areas.

A. (1) are allowed, provided that ALL Control Rods are fully inserted AND a Control Rod Withdrawal Block is applied (2) 584 B. (1) will be suspended and can NOT be continued until required refueling equipment interlocks are OPERABLE (2) 584 C. (1) are allowed, provided that ALL Control Rods are fully inserted AND a Control Rod Withdrawal Block is applied (2) 604 D. (1) will be suspended and can NOT be continued until required refueling equipment interlocks are OPERABLE (2) 604

QUESTION 99 A LOCA event has just been upgraded to a General Emergency Classification.

All Emergency response organizations were already staffed and operational when the General Emergency declaration was made.

Which ONE of the following completes both statements below in accordance with EPIP-5, General Emergency?

Notification of the NRC is required to be completed as soon as possible not to exceed (1) minutes from declaration.

The determination of protective action recommendations will come from the _(2)_.

A. (1) 15 (2) Central Emergency Control Center B. (1) 15 (2) Technical Support Center C. (1)60 (2) Central Emergency Control Center D. (1)60 (2) Technical Support Center

QUESTION 100 An ATWS has occurred on Unit 1 with the following conditions:

An Alert has been declared

The On-Call SED has assumed the duties of the SED.

There are no radiological or other hazards in the Reactor Building Subsequently:

It becomes necessary to perform 1 -EOI Appendix- 1 B, Vent and Depressurize the Scram Pilot Air Header.

Which ONE of the following completes both statements below?

The AUO will perform 1-EOI Appendix-lB at the 1)_.

In accordance with BFN-ODM-4.2, Radiological Emergency Plan (REP) Assignments, at this point in the emergency, the control room (2) allowed to DIRECTLY dispatch the AUO into this area.

A. (1) 565 elevation north east at the CRD station (2)is B. (1) CRD catwalk above Hydraulic Control Units (2)is C. (1) 565 elevation north east at the CRD station (2) is not D. (1) CRD catwalk above Hydraulic Control Units (2) is not

SRO References

16 -64

25 EOI Caution Curve 8 and Table 6)

64 2-EOI Appendix-17C, Curve 2

79 EOI Curve #3 HCTL

81 Fire Protection Report Volume 1

82 U3TS3.8.1

83 EPIP-1

87 U3TS3.3.1.i

88 Unit 1/2/3 TS 3.6.4.3, Unit 1 and2TS 3.8.i,U3 TS 3.8.2

91 Ui IS 3.3.2.1

92 NPG-SPP-03.5 Appendix A

96 U3 IS 3.5.1

Curve 3 Heat Capacity Temp Limit 250 240 -

LL.

20-U E 210 200 190- ! preSS 1 I) 0 170 gQY I H 1 Safe I t30 150.

l j

l) 14 15 16 17 18

19 Suppr P1 Lvi (ft)

ACTION REQUIRED if above curve for existing RPV press

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT FIRE PROTECTION REPORT VOLUME 1 QA Record RO R 15 R17 EFFECTIVE DATE 05/10/2007 .0< O7./20 1.3 O.../cZ/2qt.3 10/15/2013 PREPARED L.T. Stafford D.S. Kammer D.S. Kammer D.S. Kammer CHECKED R. Abbas L.T. Stafford L.T. Stafford L.T. Stafford REVIEWED R. Abbas L.T. Stafford L.T. Stafford L.T. Stafford APPROVED R.G. Jones S. Bono S. Bono S. Bono

3.8 ELECTRICAL POWER SYSTEMS 3.81 AC Sources - Operating LCO 3,8.1 The following AG electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offite transmission network and the onsite Class 1 E AC Electrical Power Distribution System;

b. Unit 3 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit I and 2 DG(s) capable of supplying the Unit 1 and 2 4.16 kV shutdown board(s) required by LCO 3.8.7, Distribution Systems Operating.

APPLICABILITY: MODES 1,2. and 3.

ACTIONS LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite Al Verity power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months circuit inoperable, from the remaining OPERABLE offsite transmission network.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND (continued)

BFN-UNIT 3 3.8-1 Amendment No. 34-2 244 December 1, 2003

BFN EMERGENCY CLASSIFICATION PROCEDURE R0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 47 OF 205 LOSS OF AC POWER Description -

5.1-U j NOTE TABLE 1 I I Loss of normal and alternate supply voltage to ALL c unit specific 4KV shutdown boards from Table 5.1 Z for greater than 15 minutes AND At least two Diesel Generators supplying power to unit specific 4KV shutdown boards listing in Table 5.1.

OPERATING CONDITION: m ALL 5.1-Al I NOTE j TABLE 1 US 5.l-A2 I I NOTE I TABLE j US Loss of voltage to ANY THREE unit specific 4KV Loss of voltage to ALL unit specific 4KV shutdown shutdown boards from Table 5.1 for greater than boards from Table 5,1 for greater than 15 minutes.

15 minutes AkIr

?%ILJ Only ONE source of power available to the m remaining board.

OPERATING CONDITION: OPERATING CONDITION:

Mode 1 or 2 or 3 Mode 4 or 5 or Defueled 5.1-S I INOTEjTABLEI US I I 1 Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes.

m m

m

t, G) m 2

(-3 OPERATING CONDITION:

Mode 1 or 2 or 3 5.1-G I INOTEITABLEI US I I I Loss of voltage to ALL unit specific 4KV shutdown 0 boards from Table 5.1 AND rn Either of the following conditions exists;

. Restoration of at least one 4KV shutdown board is NOT likely within three hours. m

. Adequate core cooling can NOT be assured.

o 0

m z

OPERATING CONDITION:

Model or2or3

EPIP-1 BFN EMERGENCY CLASSIFICATiON PROCEDURE Unft 0 EVENT CLASSIFICATION MATRIX O2O5 NOTES 5.2 250V DC power voltage below 248 votts constitutes a loss of DC power to the affected board. The vottage readings may be obtained at the 250V Shutdown Battery Board (or the 250V Plant Battery Board) that is feeding the affected board.

CURVESITABLES:

Table 5.2-U UNIT 4KV SHUTDOWN BOARD APPLICABILITY APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT I A. B, C, AND D UNIT 2 A. B, C, AND 0 UNIT 3 3A 38 3C AND 3D

BFN I EMERGENCY CLASSIFICATION PROCEDURE I EPIP-1 I Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 49 OF 205 Description 5.2-U I INOTEITABLEI US I I Unplanned loss of 250V DC control power to ALL unit specific 4KV shutdown boards from z Table 52U for greater than 15 minutes C OR Unplanned loss of 250V DC control power to unit specific 480V shutdown boards A and B for greater than 15 minutes m

OPERATING CONDITION: Z Modes 4 or 5 1 I I I I I I

m

-4 5.2-S j NOTE I TABLE I Us I I I I Loss of 250V DC power to ALL combinations (I, II. Ill, and IV) of essential systems from Table 5.2-S for greater than 15 minutes.

m m

1, C) m z

c, OPERATING CONDITION:

Mode 1 or 2 or 3 I I I I I I I I G) m z

m 1

m m

6) m z

BFN EMERGENCY CLASSIFICATION PROCEDURE II EPIP-1 Rev. 0049 j Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 51 OF 205 HAZARDS 6.0

BFN EMERGENCY CLASSIFICATION PROCEDURE R0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 53 OF 205 Description Description 6.1-U I I I I I I I Valid, unexpected increase of ANY in-plant ARM reading to 1000 mrem/hr (except TIP room).

C r

m OPERATING CONDITION:

ALL 6.1-All I 6.1-A2I I I Valid, unexpected increase of ANY in-plant ARM Control Room radiation levels greater than reading to 1000 mremlhr (except TIP room). 15 mrem/hr.

r Personnel required in the affected area(s).

-4 OPERATING CONDITION: OPERATING CONDITION:

ALL ALL I I I I Cl,

-4 m

m m

TJ 0

m z

C, I I I I I I 0

m z

m u

r m

ni 0

rn z

C,

BFN EMERGENCY CLASSIFICATION PROCEDURE I Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX j PAGE 55 OF 205 CONTROL ROOM TURBINE FAILURE EVACUATION U tion D 1 I I I 6.3-U I I I C

Turbine failure resulting in casing penetration Z OR C Significant damage to turbine or generator seals during operation.

I m

OPERATING CONDITION:

Model,or2 6.2-Al 6.3-Al I Control Room Abandonment from entry into Turbine failure resulting in visible structural 1,2, or 3-AOl-100-2 or 0-SSI-16 for ANY Unit damage to or visible penetration of ANY of the Control Room. following structures from missles:

Reactor Building Diesel Generator Building Intake Structure Control Bay m

3 OPERATING CONDITION

OPERATING CONDITION: Mode 1 or 2 ALL 6.2S I I I Control Room Abandonment from entry into 1, 2, or 3-AOl-100-2 or 0-SSl16 for ANY Unit Control Room m AND m Control of reactor water level, reactor pressure, and reactor power (for Modes 1. or 2, or 3) or decay heat removal (for Modes 4, or 5) per G) 1, 2, or 3-AOl-i 00-2 or O-SSI-16 as applicable, can NOT be established within 20 minutes after evacuation is initiated, o OPERATING CONDITION:

ALL I I I I I 0

171 z

m x,

r m

m 0

m z

C)

BFN I EMERGENCY CLASSIFICATION PROCEDURE I EP1P-1 Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 56 OF 205 NOTES CURVESITABLES:

Table 6.4-UI APPLICABLE PLANT AREA orBuildin Refuel Floor .

r 4KV Shutdown Board Rooms 4KV Shutdown Battery Board Rooms 480V Shutdown Board Rooms RMOV Board 3A and 3B Rooms 4KV Bus Tie Board Room Control Bay Elevation 593, 606, And 617 Diesel Generator Buildings (All Elevations) neBuiIdinAllEieva ions) ntakePumjcng ste Buildin (All Elevations)

Cable TunJintake To Turbine Buildingj_

StandbyGasTreatmentdin Table 6 4-A APPLICABLE PLANT AREA Reactor Building Refuel Floor 4KV Shutdown Board Rooms 480V Shutdown Board Rooms RMOV Board 3A and 3B Rooms 4KV Bus Tie Board Room Control Bay Elevation 593, 606, And 617 Diesel Generator Buildhgs llEIev2s intake Pumping Station (All Elevations) .

Turbine 8uildinL Standby GasT

I EPIP-1 T

I BFN Unit 0 EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX J Rev. 0049 PAGE 57 OF 205 Description Description 6.4-UI I I I TABLE I 6A-U2 I Confirmed fire in ANY plant area listed in Unanticipated explosion within the protected area Table 64-UI resulting in visible damage to ANY permanent AND structure or equipment.

NOT extinguished within 15 minutes, m

OPERATING CONDITION: OPERATING CONDITION:

ALL ALL 6.4-Al ITABLEI I Fire or explosion in ANY plant area listed in Table 6.4-A affecting safety system performance OR Fire or explosion causing visible damage to permanent structure of safety systems in ANY plant area listed in Table 6.4-A.

OPERATING CONDITION:

ALL p I l I Cl)

-4 m

m m

o G) m z

o I

0 m

z m

3 I

Ii m

0 m

z C,

BFN EMERGENCY CLASSIFICATION PROCEDURE Rev 0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 58 OF 205 NOTES CURVES/TABLES:

Table 6.5166 APPLICABLE PLANT AREA Reactor Buildg__

Refuel Floor Control Bay Diesel Generator Buildings Turbine Building

Intake Pumping Station Radwaste Building I Cable Tunnel (Intake To Turbine Building)

Standby Gas Treatment Bin

I BFN Unit 0 I EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX II EPIP 1 Rev. 0049 PAGE 59 OF 205 TOXIC GASES Uescripuon 6.5-U I I I TABLEI I EITHER of the following conditions exists:

. Normal operations impeded due to access restrictions caused by toxic gas concentrations within Z any building or structure listed in Table 6,5/6.6.

. Confirmed report by local, county, or state officials that a large offsite toxic gas release has c occurred within one mile of the site with potential to enter the site boundary in concentrations at or above the Permissible Exposure Limit (PEL) causing an evacuation of any site personnel.

OPERATING CONDITION: m ALL 6.5-A I I I TABLEI I ALL of the following conditions exist:

. Plant personnel report toxic gas within any building cw structure listed in Table 6.5/6.6.

. Plant personnel report severe adverse health reactions due to toxic gas (i.e., burning eyes, throat, or dizziness), or sampling results by Fire Protection or Industrial Safety personnel indicate levels above the Permissible Exposure Limit (PEL).

. Determination by the Site Emergency Director that plant personnel would be unable to perform actions necessary to establish and maintain cold shutdown conditions while utilizing appropriate personnel protective equipment.

OPERATING CONDITION:

ALL I I I I C,)

I m

m m

u C) m z

C) 1 I I I C) m z

m

t3 m

m C) m z

C,

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit C EVENT CLASSIFICATION MATRIX PA 2

0 O5 Description 6.7U 1 I I I I I I 1 A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as reported by Z the Security Shift Supervisor.

OR 2 A credible Browns Ferry threat notification OR

3. A validated notification from NRC providing information of an aircraft threat. m z

OPERATING CONDITION:

ALL 67-Al I I 1 I I I I 1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA as reported by the Security Shift Supervisor.

m

2. A validated notification from NRC of an airliner attack threat within 30 minutes of the site.

OPERATING CONDITION:

ALL 6.7S 1 I I I I I I cI A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the m Security Shift Supervisor 0

OPERATING CONDITION: m ALL 6.7G I I I I 1 C)

1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate ra equipment required to maintain safety functions. r ru OR

2. A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT z fuel damage is likely for a freshly off-loaded reactor core in pool.

OPERATING CONDITION:

ALL

BFN I EMERGENCY CLASSIFICATION PROCEDURE I

EPIP-1 Rev. 0049

[ Unit 0 EVENT CLASSIFICATION MATRIX PAGE 65 OF 205 VEHICLE CRASH Description 6.8-UI I I I I Vehicle crash (for example; aircraft or barge) into plant structures or systems within the protected area boundary.

C I

in OPERATING CONDITION: in ALL 6.8-Aj I I I Vehicle crash (for example aircraft or barge) into ANY plant vital area f

m

u

-1 OPERATING CONDITION:

ALL I

C,)

-I m

in in

3 0

in z

C>

I 0

m z

m

v r

m m

C) m z

C,

I BFN EMERGENCY CLASSiFICATION PROCEDURE I

EPIP1 Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 67 OF 205 SPENT FUEL STORAGE Description 6.9-UI I I I I Damage to a loaded cask CONFINEMENT BOUNDARY from ANY of the following: C

. Natural phenomena (e.g.. seismic event, tornado, flood, lightning, snow/ice accumulation, etc.>

Accident (e.g.. dropped cask, tipped over cask. explosion, missile damage, lire damage, burial under C debris, etc.).

. Judgement of the Site Emergency Director that the CONFiNEMENT BOUNDARY damage is a m degradation in the level of safety of the ISFSI.

m OPERATING CONDITION:

ALL a I I I I

x I

m z

I I I I I C,,

m m

m u

C) m z

C) 1 I Ii C) m z

m I

m m

0 C) m z

C,

BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 EVENT CLASSIFICATION MATRIX PAGE 69 OF 205 NATURAL EVENTS 7.0

I BFN I EMERGENCY CLASSIFICATION PROCEDURE j EPIP-1 Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 71 OF 205 EARTHQUAKE Uescrlphon 7.1-U 1 1 1 1 Valid annunciation in Unit 1 Control Room, Panel 1-XA-55-22C, Window 5, START OF STRONG MOTION ACCELEROGRAPH z C

C,)

MILJ Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

OPERATING CONDITION: rn ALL 7.1Al I Valid annunciation in the Unit 1 Control Room, Panel 1-XA-5522C, Window 6, 12 SSE RESPONSE SPECTRUM EXCEEDED AND Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

-4 OPERATING CONDITION:

ALL I I 1 1 C,)

4 m

m m

J G) m z

C)

I I 0

m z

m 0

r m

m 0

m z

C,

I 8FN I EMERGENCY CLASSIFICATION PROCEDURE I EPIP-1 Unit 0 EVENT CLASSIFICATION MATRIX I Rev. 0049 I PAGE 73 OF 205 TORNADO I HIGH WINDS I Description 72-UI 1 I Report by plant personnel of tornado striking within the protected area boundary.

C Cl)

C m

OPERATING CONDITION:

ALL 72Aj I I I i Tornado striking plant vital area OR Onsite wind speed above 90 MPH as indicated using the meteorological data screen of the Integrated Computer System (ICS). I OPERATING CONDITION:

ALL I

C,)

-1 m

m m

C) m z

C-)

II C) m z

m I

m m

C)

I,,

z C-)

I BFN 1 EMERGENCY CLASSIFICATION PROCEDURE I EPIP-1 Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 75 OF 205 FLOOD Description 7.3-UI 1 I I Wheeler Lake level exceeds or is predicted to exceed elevation 565 feet.

z AND c:

Water entering permanent plant structures due to flooding.

m OPERATING CONDITION: m ALL 7.3-Al I I I I Wheeler Lake level exceeds or is predicted to exceed elevation 565 feet.

AND I-EITHER of the following conditions exists: m

. Breech or failure of any water-tight structure is causing flooding of the structure

. Equipment required for safe shutdown is affected.

OPERATING CONDITION:

ALL I I I I C,,

-4 m

m m

0 C) m z

C)

I I I I C) m z

m 0

r m

F C)

F, z

C-)

I BFN Unit 0 EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX R0049 PAGE 77 OF 205 EMERGENCY DIRECTOR JUDGMENT 8.0

BFN EMERGENCY CLASSIFICATION PROCEDURE Rv0049 Untt 0 EVENT CLASSIFICATION MATRIX PAGE 79 OF 205 TECHNICAL SPECIFICATIONS uescraptian 8.1-U I I I I lnabhty to reach required shutdown condition (Mode 3 or Mode 4) within Z Technical Specification Limiting Conditions for Operation (LCO) limits.

C r

m OPERATING CONDITION: ITI Modelor2or3 r

m v

I Co

(

m m

m C) m z

0 I

C) m z

m I

m m

3 C) m z

C)

EPIP-1

BFN EMERGENCY CLASSIFICATION PROCEDURE

Unit 0 EVENT CLASSIFICATION MATRIX NOTES CIJRVESITABLES:

Table 8.2-U LOSS OF COMMUNICATIONS Onsite Communications Offsite Communication Plant Phone System Node 1 Bell Unes Two-Way Radio System Digital Microwave (N SS Sound Power Phones çer enc Telecommunication System NextelComm unication Cellular Phones (If Available)

Health Physics Radio Network

BFN EMERGENCY CLASSIFICATION PROCEDURE R0049 UndO EVENT CLASSIFICATION MATRIX PAGE 81 OF 295 LOSS OF COMMUNICATION Description

&2-U I I ITABLE I I Unplanned loss of onsite communication listed in Table 82-U that defeats the Plant Operations Staffs C ability to perform routine operations I,

OR Unplanned loss of ALL off-site communication listed in Table 8.2-U, m

OPERATING CONDITOIN:

ALL I

I m

0 Ii C,)

4 m

m m

C) m z

C, I

C) m z

ru 0

I m

Ill z

C) ru z

C)

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 EVENT CLASSIFICATION MATRIX PAGE 82 OF 205 NOTES 8.3 Significant Transient is an unplanned event involving one or more of the following:

(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.

CURVESITABLES:

Table &3-S APPUCABLE SAFETY FUNCTIONS Reactor Power Reactor Pressure Reactor Level Subcriticality Drywell Temature L,pfessure yession Chamber Pressure ppsionf!ol Level

F BFN I EMERGENCY CLASSIFICATION PROCEDURE I EPIP-1 I Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 83 OF 205 LOSS OF ASSESSMENT CAPABILITY Llescnptlon 8.3-UI I I I I Unplanned loss of most or all safety system annunciators or indicators which causes a significant loss of plant assessment capability for greater than 15 minutes z AND C Compensatory non-alarming safety system indications are available (SPDS. ICS)

AND in the opinion of the Shift Manager, increased surveillance is required to safely operate the plant, OPERATING CONDITION:

MODE1or2,or3 8.3-A INOTEI I Unplanned loss of most or all safety system annunciators or indicators which causes a significant loss of plant assessment capability for greater than 15 minutes AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant AND EITHER of the following conditions exists: m Compensatory non-alarming safety system indications are NOT available (SPDS, ICS)

A significant transient is in progress.

OPERATING CONDITION:

MODE tor2,or3 8.3-S I I NOTE I TABLEI I Loss of most or all annunciators associated with safety systems AND Compensatory non-alarming safety system indications are NOT available (SPDS, lOS) m AND rn Indications needed to monitor safety functions are NOT available (Refer to Table 8.3-S)

AND A significant transient is in progress.

OPERATING CONDITION:

MODE1.or2,or3 -<

I I I I I C) rn z

m xl

t m

m xl C) m z

C-)

BFN I EMERGENCY CLASSIFICATION PROCEDURE I Rev. 0049 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 84 OF 205 NOTES 8.4-U Table 8.4-U contains only example events that may justify Unusual Event classification. This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but warrant declaration of an emergency because conditions exists which the Emergency Director believes to fall under the Unusual Event Classification, Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.4-A This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the Alert classification Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.4-S This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the Site Area Emergency classification. Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.4-G This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the General Emergency classification. Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

CURVESITABLES:

Table 8,4-U OTHER EXAMPLE UNUSUAL EVENTS jjjransienes onse Unexpected Or Not Understood Unanalyzed Safety System Configuration Affecting, Threatening Safe Shutdown Inadequate Personnel To Achieve Or Maintain Safe Shutdown Degraded Plant Conditions Beyond License Basis Threatening Safe Operation Or Safe Shutdown Emergency Procedures Not Adequate To Maintain Safe Operation Or Achieve Safe Shutdown

l EPIP-1 J BFN I EMERGENCY CLASSIFICATION PROCEDURE I Rev. 0049 J Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 85 OF 205 OTHER I Uescriptton

&4-U I I NOTE I TABLEI I Events are in process or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive C material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs, Refer to Table 8.4U for examples.

C OR in Any loss or any potential loss of containment.

OPERATING CONDITION:

ALL 8.4-A I INOTEI I I Events are in process or have occurred which involve an actual or potential substantial degradation in the level of safety of the plant or a security event that involves probable 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.

OR Any loss or potential loss of fuel cladding or RCS pressure boundary.

OPERATING CONDITION:

ALL BA-S I I NOTE I I I 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) prevent effective access to equipment needed for 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.

OR Ifl Any loss or potential loss of both fuel cladding and RCS pressure boundary.

OR m Potential loss of either fuel cladding or RCS pressure boundary and loss of any additional barrier.

OPERATING CONDITION:

ALL 8.4-GI INOTEI I I 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 offsite for more than the immediate site area.

OR Loss of any two barriers and potential loss of third barrier.

0 rn z

C.)

OPERATING CONDITION:

ALL

RPS Instrumentation 3.311 3.3 INSTRUMENTATION 3.3.11 Reactor Protection System (RPS) Instrumentation LCO 3.3,11 The RPS instrumentation for each Function in Table 3.3.11-1 shall be OPERABLE.

APPlICABILiTY: According to Table 3.311-1.

ACTIONS

-NOTE--

Separate Condition entry is allowed for each channel CONDITION REQUIRED ACTION O1PLETION TIME A. One or more required Al Place channel in trIp. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR A.2 -NOTE---

Not applicable for Functions 2.a, 2.b, 2,c, Zd, or Zr.

Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

(continued)

BEN-UNIT 3 3.3-1 Amendment No. 212, 213,221 September 27 1999

RPS Instrumentation 3.311 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. -NOTE-- B. I 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,or2,f, .QE 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 Cl 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,

0. Required Action and 0.1 Enter the Condition Immediately associated Completion referenced in Time of condition A B or TableS 3 1 1-1 for the C not met. channel.

E. As required by Required El Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action 0 1 and POWER to < 30% RTP referenced in Table 3.3.1.1-1.

F. As required by Required Fl Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action Dl and referenced in Table 3,31.11.

(continued)

BFN-UNIT 3 3.3-2 Amendment No. 212. 2i3 221 September27. 1999

RPS InstrumentatIon 3311 ACTIONS (continued)

CONDITION REQUIRED ACTiON COMPLETION TIME (3. As required by Required 01 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action DI and referenced in Table 33i 11.

H. As required by Required HI Initiate action to fully Immediately Action 01 and insert all insertable referenced in control rods in core cells Table 3.31Ii. containing one or more fuel assernbhe&

I. As required by Required 11 Initiate alternate method 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D. 1 and to detect and suppress referenced in Table thermal hydraulic

33. 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 3 33-3 Amendment No, 212,213, 221.231 September 13, 2001

RPS Instrumentation 33i1 Tab1e331i-1 rpai f3 Ra Peactn Syen ffistrunewtHn APPUCABLE CONDmONS MODES OR REOURED REFERENCED FUNCTiON OTHER CHANNELS FROM SURVELL4NCE ALLOWABLE SPECIRED PER TRiP REO1MRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION DJ

2. kg Por Rae Montoes a NeoFki High. SR 33,1ti (Se1dQw) SR &3. i16 SR S3t1.7 SR a3t13 SR 33i1I

b. FIcBIasdSicuiatd I F SR33Ii1 TbomaI Power HIGh SF 33. 112 SR 331t7 SR 33L113 SR 331118 c NtnFux - Hi 1 F SR 33111 SR 3at12 SR 13112 SR 331113 SR 131116

{b) Ead APRM &arneI ptovid inputto both trip

RPS Instrumentation 331i Tabte 3.3t14 pae 3 d3 Rea Protecten System stenzbn APPU SE TN MOD P U E P FEPEN ED FUNCTION ER CH N.A. PP U ELLA E LOWABLE P WE FE PQ RED P I E AU TON ATI

2. Average Power Range Morcra (onbnued)

d. nop 12 6 SR 31J1 NA e, 2-Out-0f4 Vster 12 2 6 SR 33ii1 SR 33ttI4 SR 33,1tI

f. 0RPM Upscale SR S31ii NA SR 3:al:I.7 SR 33L11 SR 331iIe SR 3:3iii7 Ead FPRM dven& pd sipsts tbh stems.

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LOG 3.6.4.3 Three SGT subsystems shaH be OPERABLE.

APPLICABILITY: MODES 1,2. and 3, Durhg operations with a potential for draining the reactor vessel (OPD RV5).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not NP met in MODE 1,2, or3.

B.2 Be in MODE 4, 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

BFN-UNIT 1 3.6-51 Amendment No. 34. 251 September 27, 2004

SGT System 3643 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TI ME C. Required Action and C.1 Place two OPERABLE Immediately associated Completion SGT subsystems in Time of Condition A not operation.

met during OPDRVs.

OR C2 Initiate action to suspend Immediately OPDRVs.

D. Two or three SGT Di Enter LCO 3O3. Immediately subsystems inoperable in MODE 1. 2. or 3.

(continued)

BFN-UNIT 1 3.652 Amendment No. 234, 251 September 27, 2004

SGT System 3643 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two or three SGT El Initiate action to suspend Immediately subsystems inoperabie OPDRVs, during OPDRVs.

BFN-UNIT 1 3653 Amendment No, 234, 251 September 27, 2004

SGT System 3&4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas TreaUTent (SGT) System LCO 3.6.4.3 Three SOT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, Dunrig operations with a potential for draining the reactor vessel OPDRVs),

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SOT subsystem A.! Restore SOT subsystem 7 days inoperable, to OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion lime of Condition A not N2 met in MODE 1,2, or 3.

8.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continuec1 BFN-UNIT 2 3.6-51 Amendment No. 2, 290 September 27, 2004

SGT System 3,&4.3 ACTIONS (continued)

CONDITION REQUIRED COMPLETION TIME C, Required Action and C. 1 Place two OPERABLE Immediately associated Completion SGT subsystems in nine of Condition A not operation.

met during OPDRVs, OR C.2 Initiate action to suspend Immediately OPDRVs.

0. Two or three SGT Di Enter LCO 3.0.3. immediately subsystems inoperable in MODE 1. 2, or 3.

(continued)

BEN-UNiT 2 3.6-52 Amendment No. 2., 290 September 27, 2004

SGT System 3643 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E Two or three SGT El Initiate action to suspend Immediately subsystems inoperable OPDRVs.

during OPDRVS.

8FNUNiT 2 36-53 Amendment No. 253,290 September 27, 2004

SGT System 36,4,3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SOT) System LCO 3.6.4.3 Three SOT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, During operations with a potential for draining the reactor vessel OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A. I Restore SGT subsystem 7 days inoperable, to OPERABLE status.

B. Required Action and B. 1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not ,NP.

met in MODE 1,2,013.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

BFN.UNIT 3 3.651 Amendment No. 2i2. 249 September 27. 2004

SGT System 36.43 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C Required Action and Ci Place two OPERABLE ImmedateIy associated completion SGT subsystems in Time of Condition A not operation.

met during OPDRVs.

OR C.2 Initiate action to suspend Immediately OPDRVs.

0. Two or three SOT Di Enter LCO 3O3. Immediately substems inoperable in MODE i,2or3.

(continued)

BFN-UNIT 3 36-52 Amendment No 2I. 249 September 27, 2004

SGT System 36A.3 ACTIONS (conUnued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two or three SGT E.1 Initiate action to suspend Immediate subsystems inoperable OPDRVs.

during OPDRVs.

8FN-UNIT 3 3653 Amendment No 242-. 249 September 27, 2004

3,8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8,1 The following AC electric& power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class I E AC Electrical Power Distribution System;

b. Unit 1 and 2 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit 3 DG(s) capable of supplying the Unit 3 4.16 kV shutdown board(s) required by LCO 3.8.7. Distribution Systems -

Operating.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite Al Verify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months circuit inoperable, from the remaining OPERABLE offsite AND transmission network.

Once per S hours thereafter AND (continued)

BFNUNIT I 3.8-1 Amendment No. 234 249 December 1, 2003

ACTIONS CONDITION REQUI RED ACTION COMPLETION TIME A. (continued) A2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no offsite discovery of no power available offsite power to inoperable when the one shutdown redundant required board concurrent feature(s) are inoperable, with inoperabNity of redundant required feature(s)

AND A.3 Restore required offsite 7 days circuit to OPERABLE status. AND 21 days from discovery of failure to meet LCO B. One required Unit I and 2 B.1 Verify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months DS inoperable, from the offsite transmission network. AND Once per S hours thereafter AND (continued)

BFN-UNIT 1 3,8-2 Amendment No. 280

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued> 8.2 Evaluate availability of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months both temporary diesel generators (TOGS).

AND AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter 8.3 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s), supported by discovery of the inoperable Unit I and Condition B 2 DG, inoperable when concurrent with the redundant required inoperability of feature(s) are inoperable. redundant required feature(s)

AND 8.4.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Unit 1 and 2 DS(s) are not inoperable due to common cause failure.

OR 8.4.2 Perform SR 3.8.1.1 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE Unit I and 2 DG(s).

AND (continued)

BFNUNIT 1 3.83 Amendment No. 280

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B5 Restore Unit 1 and 2 DG 7 days from to OPERABLE status. discovery of unavaiIabity of TDG(s)

AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition B entry

> 6 days concurrent with unavailability of TDG(s)

AND 14 days AND 21 days from discovery of failure to meet LCO (continued)

BFN-UNIT I 383a Amendment No. 280

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One division of 480 V Cl Restore required division 7 days load shed logic of 480 V load shed logic inoperable, to OPERABLE status.

D. One division of common Dl Restore required division 7 days accident signal logic of common accident inoperable, signal logic to OPERABLE status.

E. Two required offsite El Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable, feature(s) inoperable discovery of when the redundant Condition E required feature(s) are concurrent with inoperable. inoperability of redundant required feature(s)

AND E.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

(continued)

BEN-UNIT 1 38-4 Amendment No. 24

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

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

Only applicable when more Enter applicable Conditions and than one 4.16 kV shutdown Required Actions of LCO 3.8.7.

board is affected, Distribution Systems -

- Operating when Condition F is entered with no AC power source F. One required offsite to any 4.16 kV shutdown board.

circuit inoperable.

AND P.1 Restore required offsite 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months circuit to OPERABLE One Unit 1 and 2 DG status, inoperable.

OR F.2 Restore Unit 1 and 2 DG 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to OPERABLE status.

NOTE Applicable when only one 4.16 kV shutdown board is affected.

G. One required offaite G. I Declare the affected immediately circuit inoperable. 416 kV shutdown board inoperable.

AND One Unit 1 and 2 DG inoperable.

(continued)

BFN-UNIT 1 38-5 Amendment No, 234

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Two or more Unit 1 H.1 Restore all but one Unit 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 2 DGs and 2 DG to OPERABLE inoperable, status.

Required Action and LI Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Associated Completion Time of NP Condition A. B. C, D, E F or H not met I2 Be n MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months J. One or more required J.1 Enter LCO 10.3. Immediately offsite circuits and two or more Unit 1 and 2 DGs inoperable.

OR Two required offsite circuits and one or more Unit 1 and 2 DGs inoperable.

OR Two divisions of 480 V load shed logic inoperable.

OR Two divisions of common accident signal logic inoperable.

(continued>

BFN-UNIT 1 3.86 Amendment No. 234

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME K. One or more required KI Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from Unit 3 DSs feature(s) supported by discovery of inoperable, the inoperable Unit 3 OG Condition K inoperable when the concurrent with redundant required inoperability of feature(s) are inoperable, redundant required feature(s)

AND K.2 Declare affected SGT and 30 days CREV8 subsystem(s) inoperable.

BFNUNIT 1 38-7 Amendment No. 234

3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:

a, Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System:

b. Unit 1 and 2 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit 3 DG(s) capable of supplying the Unit 34.16 kV shutdown board(s) required by LCO 3.8.7, Distribution Systems -

Operating.

APPLiCABILITY: MODES 1.2, and 3.

ACTIONS UCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Verify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months circuit inoperable, from the remaining OPERABLE offsite AND transmission network.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND (continued)

BFN-UNIT 2 3.81 Amendment No. 253 286 December 1. 2003

ACTIONS CONDITION REQUIRED ACTION COMPLETION TI ME A. (continued) A2 Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no offsite discovery of no power available offsite power to inoperable when the one shutdown redundant required board concurrent feature(s) are inoperable, with inoperability of redundant required feature(s)

AND A.3 Restore required offsite 7 days circuit to OPERABLE status. AND 21 days from discovery of failure to meet LCO B. One required Unit 1 and 2 B.1 Verify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months DG inoperable. from the offsite transmission network. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND (continued)

BFN-tJNIT 2 3.8-2 Amendment No. 307

ACTIONS (contInued I CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Evaluate availability of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months both temporary diesel generators (TDGs).

AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter 8,3. Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s), supported by discovery of the inoperable Unit I and Condition B 2 DG, inoperable when concurrent with the redundant required inoperabillty of feature(s) are inoperable, redundant required feature(s)

AND 8,4.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Unit I and 2 DG(s) are not inoperable due to common cause failure.

OR B.4.2 Perform SR 3.8.1.1 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE Unit I and 2 DG(s).

AND (continued) I BFN-UNIT 2 3.8-3 Amendment No. 307

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B5 Restore Unit 1 and 2 DG 7 days from to OPERABLE status. discovery of unavaflability of TDS(s)

AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition B entry 6 days concurrent with unavailability of TDG(s)

AND l4days AND 21 days from discovery of failure to meet LCO (continued)

BFNUNIT 2 3.83a Amendment No. 307

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One division of 480 V Ci Restore required division 7 days load shed logic of 480 V load shed logic inoperable, to OPERABLE status.

D. One division of common Di Restore required division 7 days accident signal logic of common accident inoperable, signal logic to OPERABLE status, E. Two required offsite El Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable, feature(s) inoperable discovery of when the redundant Condition E required feature(s) are concurrent with inoperable. inoperability of redundant required featurets)

AND E2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

(continued)

BFN-UNIT 2 38-4 Amendment No, 253

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

NOTE-- -------NOTE-Only applicable when more Enter applicable Conditions and than one 4.16 kV shutdown Required Actions of LCO 3.8.7.

board is affected, Distribution Systems

Operating, when Condition F is entered with no AC power source F. One required offsite to any 4.16 kV shutdown board.

circuit inoperable.

AND F.1 Restore required offsite 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months circuit to OPERABLE One Unit I and 2 DG status.

inoperable.

OR F.2 Restore Unit 1 and 2 DG 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to OPERABLE status.

NOTE Applicable when only one 4.16 kV shutdown board is affected, G. One required offsite G.1 Declare the affected Immediately circuit inoperable. 4.16 kV shutdown board in operable.

AND One Unit I and 2 DG inoperable.

(continued)

BFN-UNIT 2 3.85 Amendment No. 253

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Two or more Unit I Hi Restore all but one Unit 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 2 DGs and 2 DG to OPERABLE inoperable, status.

Required Action and LI Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Associated Completion Time of AND Condition A, B, C, D, E F or H not met I2 Be in MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months J. One or more required JI Enter LCO 3.0.3. Immediately offsite circuits and two or more Unit I and 2 DGs inoperable.

OR Two required offsite circuits and one or more Unit I and 2 DGs inoperable.

OR Two divisions of 480 V load shed logic inoperable.

OR Two divisions of common accident signal logic inoperable.

(continued)

BFN-UNIT 2 3.8-6 Amendment No. 253

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TI ME K. One or more required 1(1 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from Unit 3 DGs feature(s) supported by discovery of inoperable, the inoperable Unit 3 DG Condition K inoperable when the concurrent with redundant required inoperability of feature(s) are inoperable, redundant required featurefs)

AND K.2 Declare affected SGT and 30 days CREVs subsystem(s) inoperable.

3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit connected between the offsite transmission network and the onsite Class 1 E AC electrical power distribution subsystem(s) required by LCO 3.8.8. Distribution Systems Shutdown; b, Two of the four Unit 3 diesel generators (DGs) each capable of supplying one 4.16 kV shutdown board of the onsite Class I E AC electrical power distribution subsystem(s) required by LCO 3,8,8, Distribution Systems Shutdown: and

c. Unit 1 and 2 DGs capable of supplying the Unit 1 and 24.16kV shutdown boards required by LCO 3,8.8.

APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment.

BFN-UNIT 3 3.8-14 Amendment No. 212

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite ---------------NOTE--

circuit inoperable. Enter applicable Condition and Required Actions of LCO 3.8.8.

with any required 4.16 kV shutdown board not energized from a qualified source as a result of Condition A.

A.1 Declare affected required Immediately feature(s) with no qualified offsite power available inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment, AND (continued) 8FN-UNIT 3 3.8-15 Amendment No. 212

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2,3 Initiate action to suspend Immediately operations with a potential for draining the reactor vessel (OPDRVs),

AND A,2.4 Initiate action to restore immediately required offsite power circuit to OPERABLE status.

B. One or more required B1 .1 Suspend CORE immediately Unit 3 DGs inoperable. ALTERATIONS.

AND B,1 .2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.

AND 8.1.3 Initiate action to suspend immediately OPDRVs.

AND B. 1.4 Initiate action to restore immediately required Unit 3 DGs to OPERABLE status.

(continued)

BFN-UNIT 3 3,8-16 Amendment No. 212

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more required Ci Declare affected SGT and 30 days Unit 1 and 2 DGs CREV subsystem(s) inoperable. inoperable.

Immediately from discovery of Condition C concurrent with inoperability of redundant required feature(s)

BFN-UNIT 3 3.8-17 Amendment No. 212

33 INSTRUMENTATION 3321 Control Rod Block Instrumentation LCO 332. I The control rod block instrumentation for each Function in Table 332l-i shall be OPERABLE.

APPLICABILITY: According to Table 332l-1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A, One rod block monitor Al Restore RBM channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (RBM) channel OPERABLE status.

inoperable B. Required Action and BI Place one RBM channel 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated completion in trip.

Time of Condition A not met.

OR Two RBM channels inoperable.

C Rod worth mrnimizer C I Suspend control rod lmmedately (RWM) inoperable during movement except by reactor startup. scram.

OR (continued)

BFN-UNIT I 33-I5 Amendment No. 234

ACTIONS CONDITION REQU IRED ACTION COMPLETION TIME C. (continued) C.2ii Verify 12 rods Immediately withdrawn OR C2t2 Verify by administrative Immediately methods that startup with RWM inoperable has not been performed in the last calendar year.

AND C.22 Verify movement of During control rod control rods is in movement compliance with banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.

D. RWM inoperable during D. 1 Verify movement of During control rod reactor shutdown, control rods is in movement compliance with BPWS by a second licensed operator or other qualified member of the technical staff, (continued)

BFN-UNIT 1 31-16 Amendment No. 234

ACTIONS (continued)

CONDITION REQUIRED ACTION COW PLETION TIME E. One or more Reactor El Suspend control rod Immediately Mode Switch Shutdown

- withdrawal.

Position channels inoperable.

AND E2 Initiate action to fully Immediately insert all insertable control rods in core cells containing one or more fuel assemblies.

BFN-UNIT I 33-17 Amendment No 234

Taea3.2,1I (page lot 1 Ccwcd Pod Elcdl nstrumentaon APPUCABLE MODES OR FUNCVON OTHER REQUIRED SURJEtJANCE 4..LOWABLE SPECRIED CHANNELS REQUIREMENTS VALUE CONDITIONS 1 Rod EIod Monltor

a. LPorRangeUpscale (a) 2 SR 3.3.2.1.1 (e)

SR 3.3.2.1.4 SR 3.3.21.8

b. lntermedlate Porier Range Upscale (b) 2 SF1 3.3.2.1.1 (e(

SR 3,3.2.1.4 SR 3.3.2.1.9

c. HighPrRgeUpscale 1 (f)1g 2 SR 3.3.21.1

. SR 3.3,2.1.4 SR 3.3.2.1.8

d. mop (g),(hl 2 SR 3.3.2.1.1 NA

e. Doenscale (g),(h) 2 SR3.3.2.1.i (I)

SR 3.3.2.IA

2. RcWathMlrmlzer it°)(°J 1 SR 3.3.21.2 NA

. SR 3.3.2.1.3 SR 32.2.1.6 SR 53.2.1.7

.3. Reactor Mode Switch

Shudoen PosItion- (It) 3 SR 3.3.2.1.6 NA (a THERMAL POWER 27% and 83% RI? and MCPR less than th value specl6ed in the COLR.

Ib) THERMAL POWER> 62% and 93% RI? and MCPR less than the value specied In the CCILR.

(0) With THERMAL POWER 10% RIP, excp during the reactor shuLbeei process Wthe coupling of each withdrawn control rod has been corthined.

(dl Reactor mode switch In the stiutooen- position.

(e) Less than or equal to the Atiowable Value speci6ed in the COLR.

(f( THERMAL POWER> 8214 and ((Ill RIP and MCPR less than the value specified In the COLR.

Ig) THERMAL POWER 0% RIP and MCPR less than the value specIfied in the COLR.

Ih) THERMAL POWER 2714 and O% RTP and MCPR less than the value specIfied In the COLR Ii) Greater than or equal to the Allowable Value specified in the COLR.

BFN-UNT 1 Amendment No. 231. 2-Z, 276 November 19, 2009

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev. 0009 Processes Page 20 of 100 Appendix A (Page 1 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 1.0 PURPOSE This Appendix identifies reporting requirements; and instructions for detemilning reportability, preparation, and transmittal of LER5; and notification to NRC for events occurring at T\As licensed nuclear plants.

2.0 SCOPE TVA is required by §5012 and §50.73 to promptly report various types of conditions or events and provide written follow-up reports, as appropriate. This appendix provides reporting guidance applicable to licensed power reactors.

NOTES

1) Appendix B provides additional reporting criteria found in §Part 20, 30, 40, and 70 that may he applicable to events involving byproduct, source or special nuclear material possessed by the licensed nuclear plant. Site Licensing and Site RadCon are responsible for making the reportability determinations for §Part 20, 30, 40, or 70 events associated with their site. Corporate Licensing and Corporate RadChem are responsible for making the reportahility determinations for §Part 20, 30, 40. or 70 events associated with all other TVA licensed activities. Licensing is responsible for developing (with input from affected organizations> and submitting the immediate notification and written reports to NRC in accordance with §Part 20, 30, 40, or 70 requirements. Reporting requirements for personnel exposure required by §Part 20 are contained in RCTP-105, Personnel lnprocessing and Dosimetry Administrative Processes.

2) Appendix C contains the criteria for reporting if events or conditions affecting ISFSI.

TVA, as the general licensee of the ISESI, is required by §72.21 $ to make initial and written reports in accordance with §72.74 and §72.75. Operations is responsible for making the reportability determinations for §72.74 and §72.75 reports. For any event, condition, or issue having the potential for being reportable. contact Site Licensing for consultation and concurrence on the reportability determination. In no event shall the lack of licensing concurrence result in a failure to meet specified reporting timeframes.

Operations is responsible for making the immediate notification to NRC in accordance with §72.74. Operations is responsible for making the immediate, 4-hour, and 24-hour notifications to NRC in accordance with §72.75. Licensing is responsible for developing (with input from affected organizations) and submitting the written reports required by §72.75.

3) Reporting requirements for events or conditions affecting the physical protection of the licensed nuclear plant specified in §73.71 are contained in NSDP-1, Safeguards Event Reporting Guidelines. Responsibilities for reportability determinations and immediate notification requirements are assigned to Site Nuclear Security and Corporate Nuclear Security. Licensing is responsible for developing (with input from affected organizations) and submitting the written reports requwed by §73.71.

NPG Standard Regulatory Reporting Requirements NPG-SPP..03.5 Programs and Rev. 0003 Processes Page 21 of 100 Appendix A

<Page 2 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power P:lants 3.0 REQUIREMENTS NOTES

1) Internal management notification requirements for plant events are found in Appendix D. The Operations Shift Manager is responsible for notifying Site Operations Management and the Duty Plant Manager. The Duty Plant Manager is responsible for making the remaining internal management notifications.

2> NRC NUREG-i 022. Revision 3 and subsequent revisions should be used as guidance for determining reportability of plant events pursuant to §5012 and §5073.

A text searchable copy of NUREG1 022 is maintained on the TVA NPG Nuclear Licensing Webpage.

3) In addition to reviewing the clarifying discussion and examples associated with specific reporting criteria [eg, discussion of utilization of engineering judgment when evaluating Unanalyzed Conditions in NUREG 1022, Section 324(B)J, NUREG 1022, Section 2, Reporting Areas Warranting Special Mention, should also be reviewed. [R. 1]

31 Immediate Notification NRC TVA is required by §5072 to notify NRC immediately if certain types of events occur. This appendix contains the types of events and the allotted time in which NRC must be notified.

(Refer to Form NPG-SPP-03.5-i or NRC Form 361). Operations is responsible for making the reportability determinations for §5012 and §5073 reports. For any event, condition, or issue having the potential for being reportable, contact Site Licensing for consultation and concurrence on the reportability determination. In no event shall the lack of licensing concurrence result in a failure to meet specified reporting timeframes. Operations is responsible for making the immediate notification to NRC in accordance with §5012.

Notification is via the Emergency Notification System. If the Emergency Notification System is not operative, use a telephone, telegraph, mailgram, or facsimile.

NOTE The NRC Event Notification Worksheet may be used in preparing for notifying the NRC, This Worksheet may be obtained directly from the NRC website (www.nrc.gov) under Form 361, or WA NPG Form NPG-SPP-03,5-1 may be used.

A. The Immediate Notification Criteria of §50.72 is divided into 1-hour, 4-hour, and 8-hour phone calls, Notify the NRC Operations Center within the applicable time limit for any item which is identified in the Immediate Notification criteria.

B. The following criteria require 1-hour notification:

NPG Standard Regulatory Reporting Requirements NPG.SPP.035 Programs and Rev. 0009 Processes Page 22 of 100 Appendix A (Page 3 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 31 Immediate Notification NRC (continued)

(Technical Specifications) Safety Limits as defined by the Technical Specifications which have been violated.

2. §50.72 (a)(l)(i) The declaration of any of the Emergency classes specified in the licensees approved Emergency Plan.

NOTE If it is discovered that a condition existed which met the Emergency Plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of discovery, an ENS notification (and notification of the Operations Duty Specialist), within one hour of discovery of the undeclared (or misclassified) event, shall be made. However, actual declaration of the emergency class is not necessary in these circumstances.

3. §50.72(b).(i)) Any deviation from the plants Technical Specifications authorized pursuant to §50.54(x).

4. 10 CFR 73, Appendix S. paragraph I Safeguards Events. The requirements of

§7311. Reporting of Safeguard Events. are also applicable. Refer to NSDP- 1, Safeguards Event Reporting Guidelines, for additional information.

a. Any event in which there is reason to believe that a person has committed or caused, or attempted to commit or cause, or has made a credible threat to commit or cause:

(1) A theft or unlawful diversion of special nuclear material; or (2) Significant physical damage to a power reactor or any facility possessing SSNM or its equipment or carrier equipment transporting nuclear fuel or spent nuclear fuel, or to the nuclear fuel or spent nuclear fuel a facility or carrier possesses; or (3) Interruption of normal operation of a licensed nuclear power reactor through the unauthorized use of or tampering with its machinery, components, or controls including the security system. [Note: a Confirmed Cyber Attack at any NPG site is reported to the NRC in accordance with the requirements of 10 CFR 73, Appendix S. Review the Incident Categorization section in NPG-SPP-12.8.8.J

b. An actual entry of an unauthorized person into a protected area, material access area, controlled access area, vital area, or transport.

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Reporting of Events or Conditions Affecting Licensed Nuclear Power Pants Immediate Notification - NRC (continued)

c. Any failure, degradation, or the discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to a protected area, material access area, controlled access area, vital area. or transport for which compensatory measures have not been employed.

d. The actual or attempted introduction of contraband into a protected area, material access area, vital area, or transport.

C. The foflowing criteria require 4hour notification

1. §50.72(b)(2)(i) The initiation of any nuclear plant shutdown required by the plants Technical Specifications.

2. §50.72(b)(2)(iv)(A3 Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

3. §50.72(b)(2)(iv)(B) Any event or condition that results in actuation of the reactor protection system RPS) when the reactor is critical except when the actuation results from and is part of a pre-pianned sequence during testing or reactor operation, NOTES

1) NPG-SPP-05.14 provides additional instructions regarding addressing and informally communicating events to outside agencies involving radiological spills and leaks.

2) Routine or day-to-day communications between TVA organizations and state agencies typically do not constitute a formal notification to other government agencies that would require a report in accordance with §50.72(b)(2)(xi).

4. §5O.72(b)(2)xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials.

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Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 31 Immediate Notification NRC (continued)

D. The following criteria require 8-hour notification:

NOTE With the exception of Events or Conditions that Could Have Prevented Fulfillment of a Safety Function ENS notifications are required for any event that occurred within three years of discovery, even if the event was not ongoing at the time of discovery.

I §5O72(bX3)(ii)(A) Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

2, §5O,72(b)(3)(ii)(B) Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.

3. §5012(b)(3)(iv)(A) Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(8) [see list belowb except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

a. The systems to which the requirements of paragraph §50.72(bX3)(iv)(A) apply are:

NOTE Actuation of the RPS when the reactor is critical is also reportable under §50,72( )(2)(iv)(B) above.

(1) Reactor protection system (RPS) including: reactor scram or reactor trip.

(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

(3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

(4) ECS for boiling water reactors (SWRs) including: core spray systems:

high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

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Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification NRC (continued)

(5) BWR reactor core isolation cooling system.

(6) PWR auxiliary or emergency feedwater system.

(7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

(8) Emergency ac electrical power systems, including: Emergency diesel generators (EDGs).

NOTE For systems within scope, the inadvertent TS inoperability of a system in a required mode of applicability constitutes an event or condition for which there is no longer reasonable expectation that equipment can fulfill its safety function. Therefore, such events or conditions are reportable as an Event or Condition that Could Have Prevented Fulfillment of a Safety Function.

4. §5012(b)(3)(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition:

(B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

NOTE According to §50.72 (b)(3)(vi) events covered by §50.72(b)(3)(v) may include one or more procedural errors equipment failures andor discovery of design analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.

5. §50.72b)(3)(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment

NPG Standard Regulatory Reporting Requirements NPS-SPP-03.5 Programs and Rev, 0009 Processes Page 26 of 100 Appendix A (Page 7 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 31 Immediate Notification - NRC (continued) 6, §5072(h)(3)(xiii) Any event that results in a major loss of emergency assessment capability. offsite response capablilty, or offsite communications capability (e.g, significant portion of control room indication, emergency notification system, or offsite notification system).

E. Follow-up Notification (50J2(c))

With respect to the telephone notifications made under paragraphs (a) and (b) [50.72 (a) and §50 72 (b) respectively] of this sechon [50 72] m addbon to making the required initial notification, during the course of the event:

1. Immediately report:

(i) Any further degradation in the level of safety of the plant or other worsening plant conditions including those that require the declaration of the Emergency Classes. if such a declaration has not been previously made; or (ii) Any change from one Emergency Class to another, or (iii) A termination of the Emergency Class.

(1) Immediately report:

(i) The results of ensuing evaluations or assessments of plant conditions, (ii) The effectiveness of response or protective measures taken, and (ii) Information related to plant behavior that is not understood.

(2) Maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03,5 Programs and Rev. 0009

Processes Page 27 of 100 Appendix A (PageS of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 32 Twenty-Four Hour Notification - NRC Any violation of the requirement contained in specific operating hcense conditions, shall he reported to NRC in accordance with the license condition.

33 TwoDay Notification NRC -

§509(b) The NRC shall be notified of incomplete or inaccurate information which contains significant implications for the public health and safety or common defense and security Notification shall be provided to the administrator of the appropriate regional office within two working days of identifying the information. Licensing is responsible for determining reportability (with input from affected organizations) and notifying NRC in accordance with

§509.

3.4 Sixty-Day Verba Report

§50i3a)(2)(ivXA) requires that any event or condition that resulted in manual or automatic actuation of the specified systems be reported as a Licensee Event Report (LER [Refer to Appendix A, Section 35fl. This CER section also allows that in the case of an invalid actuation, other than actuation of the reactor protection system when the reactor is critical, an optional telephone notification may be placed to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER, A. Verbal Report Required Content:

If the verbal notification option is selected (NUREG 1022. Revision 3, Section 3%6.

System Actuation), instead of an LER, the verbal report:

I Is not considered an LER,

2. Should identify that the report is being made under §5073(a)(2)(iv)(A).

3. Should provide the following information:

a The specific train(s) and system(s) that were actuated.

b Whether each train actuation was complete or partial.

c. Whether or not the system started and functioned successfully.

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Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.4 Sixty..Day Verbal Report (continued)

NOTE Licensing will ensure that the information that is provided to NRC during the Sixty-Day Verbal Report is verified in accordance with NPG-SPP-03.1 0.

B. Verbal Report Development and Review Licensing will:

1. Develop (with input from responsible organization) the response (i.e.. report summary> to address the required input.

2. Ensure that the reporting details are approved by site vice president or his designee prior to making the verbal report.

C. Telephone Report Timeliness Operations will make the 60-day telephone report promptly after the response is approved by the site vice president or his designee.

3.5 Written Report NRC A, A report on a Safety Limit Violation shall be submitted to the NRC. the NSRB, and the Site Vice President if required by Technical Specifications.

B. Any violation of the requirements contained in the Operating license conditions in lieu of other reporting requirements requires a written follow-up report if specified in the license.

C. Reporting Radiation Injuries

1. §140.6(a) requires. as promptly as possible, submittal of a written notice [e.g.,

report] in the event of:

a. Bodily injury or property damage arising out of or in connection with the possession or use of the radioactive material at the licensees facility

[location]: or

b. In the course of transportation: or

c. In the event any radiation exposure claim is made. (Refer to RCDP-9.

Radiological and Chemistry Control Radiological Exposure Inquiries)

NPG Standard Regulatory Reporting Requirements NPGSPP-035 Programs and Rev. 0009 Processes Page 29 of 100 Appendix A (Page 10 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 35 Written Report NRC (continued)

2. The written notice shall contain particulars sufficient to identify the licensee and reasonably obtainable information with respect to time, place, and circumstances thereof, or the nature of the claim.

0. Licensee Event Reports A written report shall be prepared in accordance with §5013(a)(i) for items in the 60-day report criteria or Technical Specifications. The report shall be complete and accurate in accordance with the methods outlined in this procedure. The completed forms shall be submitted to the USNRC, Document Control Desk, Washington, DC 20555. NIJREG 1022, Revision 3, contains the instructions for completion of the LER form. Licensing is responsible for developing (with input from affected organizations) and submitting the written reports (or optional telephone reports [refer to Appendix A, Section 34J) required by §5073.

NOTE Unless otherwise specified in the reporting criteria below, an event shall be reported if it occurred within three years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.

E Report Criteria

1. §5073(a)(2)(i)A) The completion of any nuclear plant shutdown required by the plants Technical Specifications.

2. §5073(a)(2)(i)(B) Any operation or condition which was prohibited by the plants Technical Specifications, except when:

a. The Technical Specification is administrative in nature;

b. The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was petfomied, and the equipment was found to be capable of performing its specified safety functions; or

c. The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.

3. §50.73(a)(2)(i)C) Any deviation from the plants Technical Specifications authorized pursuant to §50.54(x)

NPG Standard Regulatory Reporting Requirements NPG-SPP03.5 Programs and Rev. 0009 Processes Page 30 of 100 Appendix A (Page 11 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.5 Written Report NRC (continued)

4. §50.73(a)(2)(ii)(A) Any event or condition that resulted in the condition of the nuclear power plant. including its principal safety barriers, being seriously degraded.

5. §50.73(a)(2)(iiXB) Any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety,

6. §50.73(a)(2)(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

7. §50.73(a)(2)(iv)(A) Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) [see list in Section 3.5E.8 below], except when

a. The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or

b. The actuation was invalid and (i) Occurred while the system was properly removed from service or (ii) Occurred after the safety function had been already completed.

NOTE In the case of an invalid actuation. other than actuation of the reactor protection system (RPS) when the reactor is critical, a telephone notification to the NRC Operations Center within 60 days after discovery of the event may be provided instead of submitting a written LER (50.73(a)). [Refer to Appendix A, Section 3.4J

8. §50.73(a)(2)(iv)(B) The systems to which the requirements to paragraph (a)(2)(iv)(A) of this section apply are:

a. Reactor protection system (RPS) including: reactor scram or reactor trip,

b. General containment isolation signals affecting containment isolation valves m more than one system or multiple man steam solaton valves MSIVs

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Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.5 Written Report NRC (continued)

c. Emergency core cooling systems (ECCS) for pressurized water reactors (PWR5) including: high-head, intermediate-head, and low-head irection systems and the low pressure injection function of residual (decay) heat removal systems

d. ECCS for boiling water reactors (BWR5) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

e BWR reactor core isolation cooling system.

L PWR auxiliary or emergency leedwater system.

g. Containment heat removal and depressurization systems. including containment spray and fan cooler systerns h Emergency ac electrical power systems, including: emergency diesel generators (EDGs)

Emergency service water systems that do not normally run and that serve as ultimate heat sinks

9. §50.73(a>(2)(v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

NOTE Events reported above may include one or more procedural errors, equipment failures, and/or discovery of design, analysis. fabrication, construction. and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this criterion if redundant equipment in the same system was operable and available to perform the required safety function

[5O.73(aX2)(vi)}

NPG Standard Regulatory Reporting Requirements NPG-SPP-035 Programs and Rev, 0009 Processes Page 32 of 100 Appendix A (Page 1 of IS)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 35 Written Report NRC (continued)

10. §50i3(a)(2)vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residuai heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

1 1. §50J3(a)(2)(viii)(A) Any airborne radioactivity release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in Appendix B to Part 20, table 2, column 1

12. §50J3a)(2)(viii)(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . exceeds 20 times the applicable concentrations specified in Appendix B to Part 20. table 2. column 2. at the point of ent into the receiving waters (ie, unrestricted area) for all radionuclides except tritiurn and dissolved noble gases.

13. §5073(a)(2)(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

a. Shut down the reactor and maintain it in a safe shutdown condition;

b. Remove residual heat;

c. Control the release of radioactive material; or

d. Mitigate the consequences of an accident.

NPG Standard Regulatory Reporting Requirements NPSSPP0&5 Programs and Rev, 0009 Processes Page 33 of 100 Appendix A (Page 14 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 35 Written Report NRC (continued)

NOTE Events covered above may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to this criterion if the event results from a shared dependency among trains or channels that is a natural or expected consequence of the approved plant design or normal and expected wear or degradatton f50 73(a)(2)(x)(B)]

14. §5013(a)(2)(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for The safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive release&

15, 10 CER 73, Appendix G, paragraph I If a one hour notification is made in Appendix A, section 31 ,BA of this procedure, then a written notification to the NRC is required within 60 days.

16. For reporting a defect found installed in the Plants Safety Related Equipment, Radioactive Wastes System, and Special Nuclear Material within an LER, §Part 21 NRC Reporting of Defects and Noncompliance, see Appendix G in this procedure.

17

a. WBN or SON shall record any occurrence of unusual or important environmental events. Unusual or important events are those that potentially could cause or indicate environmental impact causally related with station operation. The following are examples:

(1) Excessive bird impaction events; (2) Onsite plant or animal disease outbreaks; (3) Unusual mortality of any species protected by the Endangered Species Act of 1973; (4) Fish kills near the plant site;

NPG Standard Regulatory Reporting Requirements NPG.SPP-035 Programs and Rev, 0009 Processes Page 34 ci 100 Appendix A (Page 15 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.5 Written Report NRC (continued)

(5) Unanticipated or emergency discharges of waste water or chemical substances that exceeds the limits of, or is not authorized by, the NPDES permit and requires 24hour notification to the County or State of Tennessee; WBN n1y (6) Identification of any threatened or endangered species for which the NRC has not initiated consultation with the Federal Wildlife Service (FWS).

(7) Increase in nuisance organisms or conditions in excess of levels anticipated in station environmental impact appraisals.

b SON TS Appendix B compliance guidance is provided in the flowchart in NPG-SPPO55, Environmental Control, Appendix B.

c. WBN TS Appendix B compliance is met through the procedures referenced in NPG-SPP-0S5.

d. Once an unusual or important event has occurred, the required actions are:

(1) Refer to NPG-SPPO55, Environmental Control, Section Compliance with the NRC Appendix B to the Facility Operating License, for additional guidance.

(2) If required, SON or WBN Site Licensing shall make a written report to the NRC in accordance with the NRC Non-routine Report, TS Appendix B, Subsections 5.42, within 3D days, in the event of a reportable occurrence in which a limit specified in a relevant permit or certificate issued by another Federal, State, or local agency is exceeded.

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, ACTIONS LCO 3,O.4,b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TI ME A. One low pressure ECCS A.1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem(s) to OPERABLE status.

OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

(continued)

BFN-UNIT 3 3.51 Amendment No. 212, 229, 244 December 1. 2003

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

8. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AN met.

8.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued>

BFN-UNIT 3 3.5-la Amendment No, 244 December 1, 2003

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. HPCI System inoperable. Immediately AND C2 Restore HPCI System to 14 days OPERABLE status.

D. HPCI System inoperable. Di Restore HPCI System to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND OR Condition A entered.

D2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injection/spray subsystem to OPERABLE status.

E. One ADS valve Ei Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One ADS valve Ri Restore ADS valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

AND Condition A entered. F,2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injection/spray subsystem to OPERABLE status.

(continued)

BFN-UNIT 3 3.52 Amendment No, 232w 229 March 12, 2001

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Two or more ADS valves G,1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND OR G.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and dome pressure to associated Completion S 150 psig.

Time of Condition C, D, E, or F not met.

H, Two or more low pressure I-il Enter LCO 30.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperable.

BFN-UNIT 3 3.5-3 Amendment No, 242w 229 March 12, 2001

ML14142A161

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