Text

Initial Exam 2015-301 Final SRO Written Exam
	ML15040A581
Person / Time
Site:	Browns Ferry
Issue date:	02/09/2015
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-259/15-301, 50-260/15-301, 50-296/15-301 50-259/OL-15, 50-260/OL-15, 50-296/OL-15
	Download: ML15040A581 (163)
	v • d • e

U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: Facility/Unit: Browns Ferry Region: II Reactor Type: 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.

Applicant's Signature Results RO/SRO-Only/Total Examination Values 75 I 25 I 100 Points Applicant's Scores I -- I -- Points Applicant's Grade -- I -- I -- Percent

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Q76 Night shift turned over all three units operating at 100 percent power.

Unit 1 scrammed at 1000 due to a Generator Load Reject.

In accordance with NPG-SPP-03.5 which one ofthe following completes the statements below?

The latest time that the NRC is required to be notified of this event is_ (1) _,and_ (2) _is responsible for making this NRC notification.

[Reference Provided]

A. (1) 1400 (2)Site Licensing B. (1) 1800 (2) Site Licensing C. (1) 1400 (2) Operations D. (1) 1800 (2) Operations

Q77 Unit 3 is shut down due to an LCO 3.0.3 entry condition.

Mode 4 was entered

3D RHR pump is in Shutdown Cooling

Loop 1 RHR Shutdown Cooling is Not available

RWCU is in service

Drywell Equipment Hatch is open Subsequently plant equipment problems result in the following conditions:

RHR Pump 3D flow lowers to 3000 gpm and cannot be raised

3-TR-56-4 Point 7, 3-TE-56-8 Reactor Vessel Drain to RWCU indicates 218° F

Reactor Steam Dome Pressure is 1.5 psig Which ONE of the following completes both statements below?

At this flowrate, RHR Shutdown Cooling _ (1) _ considered in-service.

In accordance with EPIP-1, Emergency Plan Implementing Procedure,_ (2)_.

[Reference Provided]

A. (1) is (2) No EAL is exceeded B. (1) is (2) an Alert EAL is met C. (1) is Not (2) No EAL is exceeded D. (1) is Not (2) an Alert EAL is met

Q78 Unit 1, 2, and 3 are currently at 100% power.

While fuel movement was in progress in Unit 1 Spent Fuel Pool, several fuel bundles were damaged.

Subsequently:

The Field Assessment Team reported that the Site Boundary Radiation Reading reached 10 MREM/HR at 0840, and is now 12 MREM/HR and rising slowly at 0900.

Radiation levels on the refuel floor elevation 664 and the Recirc MG set area elevation 639 are above the Max Safe value.

Which one of the following completes both statements below?

Entry into 0-EOI 4, Radioactive Release Control,_ (1) _required.

Entry into 1-C-2, Emergency RPV Depressurization,_ (2) _required.

[Reference Provided]

A. (1) is (2) is Not B. (1) is (2) is C. (1) is Not (2) is Not D. (1) is Not (2) is

Q79 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 20% and lowering due to SLC injection

Reactor pressure 900 psig and stable

Suppression pool temperature 190 op and rising

Suppression pool level 16.0 ft. and slowly rising Which ONE ofthe following identifies the Minimum required action in accordance with 1-EOI-1, RPV Control and 1-EOI-2, Primary Containment Control?

[Reference Provided]

A. Lower reactor pressure, must maintain ~ 100 op /Hr B. Emergency depressurize (ED) the RPV using the safety relief valves C. Rapidly depressurize the RPV with the main turbine bypass valves D. Do Not ED, lower reactor pressure, OK to exceed 100 °F/Hr

Q80 An ATWS and a LOCA have occurred on Unit 2, resulting in the following plant conditions:

Suppression Chamber Pressure is 53 psig and rising 1 psig every 2 minutes

Drywell Temperature is 325°F and rising 1°F every 5 minutes

Suppression Pool Level is 18 feet Which ONE of the following identifies the required procedure(s) in accordance with 2-EOI-2, Primary Containment Control?

A. EOI appendix 12 Primary Containment Venting and EOI appendix 17B RHR System Operation Drywell Sprays B. EOI appendix 17B RHR System Operation Drywell Sprays ONLY C. EOI appendix 13 Emergency Venting Primary Containment ONLY D. EOI appendix 17B RHR System Operation Drywell Sprays and EOI appendix 13 Emergency Venting Primary Containment

Q81 An ATWS has occurred on Unit 2.

RPV water level was lowered in accordance with 2-C-5, Level I Power Control

SLC is injecting, SLC Tank level is 80%

MSRV s are being cycled for pressure control Subsequently the following indications and alarms are reported:

RPV water level is(-) 35 inches

Suppression Pool Temperature is 106 °F

SLC tank level is 63%

RPV water level is being restored to (+) 2 to (+) 51 inches

SRM PERIOD, (2-9-5A, Window 20)

APRM downscale lights extinguished Which ONE of the following describes the action(s) that is(are) required in accordance with 2-C-5, Level I Power Control?

A. Continue to raise RPV water level, but at a slower rate.

B. Stop raising RPV water level.

Maintain RPV water level with(-) 35 inches as the upper limit.

C. Stop and prevent all injection into the RPV (except RCIC, CRD & SLC).

Re-inject when RPV water level drops below(-) 50 inches.

D. Stop and prevent all injection into the RPV (except RCIC, CRD & SLC).

Re-inject when reactor power is< 5%.

Q82 Unit 2 is operating at 96% power when the following occurred:

2A Recirculation Pump speed started to rise rapidly

The US directed tripping the 2A Recirc pump

A reactor scram was inserted due to the power excursion Subsequently

2-TIS-1-60C temperature is rising rapidly

Numerous Turbine Building radiation alarms are received

The US then directs closing the MSIV s The following conditions currently exist:

All control rods are FULL IN

Reactor pressure is 780 psig and lowering

2-TIS-1-60C Main Steam Tunnel Temperature indicates 320 °F and rising slowly

B Main Steam Line has failed to isolate

2-RE-90-272A Drywell High Range Radiation Monitor indicates 310 R/HR

2-RE-90-273A Drywell High Range Radiation Monitor indicates 300 R/HR Which ONE of the following identifies the highest required Emergency Action Level/Designator to be declared in accordance with the Emergency Classification Procedure - EPIP-1?

[Reference Provided]

A. Notification ofUnusual Event B. Alert C. Site Area Emergency D. General Emergency

Q83 Unit 1 has scrammed with the following conditions:

19 Control Rods are at position 48

ATWS actions are complete

SLC is NOT injecting

MSIV s are open and Reactor pressure is stable on bypass valves

1-EOI-1, RPV Control was entered on low Reactor water level Subsequently, the OATC reports that Reactor Power is on range 8 of the IRMs.

Which ONE of the following identifies the required procedures for reactor power and level control?

A. 1-EOI-1, RPV Control, RCIQ and RCIL B. 1-AOI-100-1, Reactor Scram and RCIL C. 1-EOI-1, RPV Control, RCIQ and 1-C-5, Level I Power Control D. 1-AOI-100-1, Reactor Scram and 1-C-5, Level I Power Control

Q84 Unit 2 is operating at 100% power.

A leak has occurred in the RWCU Heat Exchanger Room.

At 0800 Heat Exchanger Room maximum safe temperature (220 °F) was reached.

RWCU Heat Exchanger Room temperature is rising at 1°F per minute

At 0815 The UO reports the following RPV water level readings:

2-LI-3-58A, Emergency Range reading(-) 4 inches 2-LI-3-208A, Normal Range reading 16 inches ASSUME that:

RPV water level remains constant RWCU Heat Exchanger room temperature rate of rise remains constant.

In accordance with Caution 1 which ONE ofthe following completes the statements below?

At 0815 _ (1) _level instrument(s) may be used to determine or trend Reactor Water level.

At 0840 based on the level indications given 2-EOI-C4, RPV flooding,_ (2) _required to be entered.

[Reference Provided]

A. (1) 2-LI-3-208A and 2-LI-3-58A (2) is B. (1) 2-LI-3-208A and 2-LI-3-58A (2) is Not C. (1) 2-LI-3-58A only (2) is D. (1) 2-LI-3-58A only (2) is Not

Q85 Unit 1 is operating at 100% power when the following alarms are received.

1-9-3B window 10 PRJ CONTAINMENTNITROGENPRESS HI

1-9-3B window 19 DRYWELL NORM OPERATING PRESS HIGH

1-9-3B window 23 DRYWELL PRESSURE HIGH

1-9-3B window 30 DRYWELL PRESS APPROACHING SCRAM Which one of the above alarms, if valid, requires the Shift Manager to classify an event in accordance with EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE?

A. Window 10 B. Window 19 C. Window23 D. Window30

Q86 Unit 2 is Shutdown with a cooldown in progress at a Reactor Pressure of 90 psig.

In accordance with the bases for Technical Specifications 3.4.7, Residual Heat Removal (RHR)

Shutdown Cooling System-Hot Shutdown, and 3.5.2, ECCS-Shutdown:

(1) How many RHR Shutdown Cooling Subsystems are there?

(2) If RHR Loop I is aligned for Shutdown Cooling can it be considered an Operable LPCI Subsystem?

A. Two; Yes B. Four; Yes C. Two;No D. Four; No

Q87 Unit 1 startup in accordance with 1-GOI-100-1A, UNIT STARTUP, is in progress.

The following conditions currently exist:

SRMs are reading between 700 and 1000 cps

All IRMs are on range 1 or 2 Subsequently:

At 08:00 SRM A began spiking causing a control rod block and was bypassed.

SRM A continued to spike until 08:20 and then returned to a stable reading, comparable to the other SRMs.

What is/are the MINIMUM action(s) required, if any, for that SRM to be returned to OPERABLE status?

A. Un-bypass the SRM, no further actions are required.

B. Observe the SRM for at least 15 minutes before returning the instrument to service with concurrence from System Engineering.

C. Perform Surveillance 1-SR-3.3.1.2.4, SRM System Count Rate and Signal to Noise Ratio Check.

D. Perform Surveillance 1-SR-3.3.1.2.5 & 6, SRM Functional Test with Reactor Mode Switch Not in Run.

Q88 Unit 3 has entered Mode 1 at 08:00 on June 1st when the following sequence of events occurs:

10:00 APRM Voter 1 failed its surveillance and did NOT generate an output signal to RPS.

11:00 The IMs report that a review of the surveillance indicates that the APRM Voter 4 also failed acceptance criteria and has been declared INOP.

Which of the following is the most limiting Technical Specification required actions for these conditions?

[Reference Provided]

A. Required Action A.1 OR A.2 must be performed by 22:00.

B. Required Action B.1 OR B.2 must be performed by 17:00.

C. Required Action C.1 must be performed by 12:00.

D. Required Action 0.1 must be performed by 23:00.

Q89 Unit 3 is operating at 100% power,at 09:00:00 a LOCA occurs.

09:00:30 (30 seconds after the LOCA) the following conditions exist:

RPV water level is (-) 125 inches and lowering

Drywell Pressure is 15 psig and rising

ADS BLOWDOWN AUX RELAYS ENERGIZED (3-9-3C, window 4) in alarm

ADS BLOWDOWN TIMERS INITIATED (3-9-3C, window 11) in alarm

ALL ECCS Systems are operating as expected 09:01:30:

ECCS systems are injecting and the Unit Supervisor has determined that Reactor water level can be restored and maintained above (-) 162 inches Which ONE of the following is required concerning ADS and what is the required procedure for Reactor water level control?

A. Do Not Inhibit ADS and execute C-1Altemate Level Control B. Do Not Inhibit ADS and execute EOI-1 RC/L C. Inhibit ADS and execute C-1Altemate Level Control D. Inhibit ADS and execute EOI-1 RC/L

Q90 Unit 2 is operating at 100% power when the following event occurs:

6-1-14 at 09:00 Reactor Vessel Steam Dome Pressure (2-PIS-3-22AA) fails upscale causing a half scram. All required Tech Spec actions were taken.

6-1-14 at 10:00 Drywell Pressure High (2-PIS-64-56B) fails downscale.

Which one of the following identifies the earliest time that Unit 2 is required to be in cold shutdown if2-PIS-64-56B cannot be restored to operable status or placed in the tripped condition.

(Consider ONLY Tech Spec 3.3.6.1, PCIS instrumentation requirements)

[Reference Provided]

A. 6-2-14 at 22:00 B. 6-2-14 at 23:00 C. 6-3-14 at 10:00 D. 6-3-14 at 22:00

Q91 Unit 3 is operating at 100% power with several control rods declared SLOW due to scram time testing data in accordance with Tech Spec 3 .1.4, Control Rod Scram Times.

(See attached illustration).

Subsequently, The CRD pump tripped and was restarted in accordance with 3-AOI-85-3, CRD System Failure.

During the time the CRD pump was not running, the CONTROL ROD DRIVE UNIT TEMP HIGH (3-9-5A, Window 17) annunciator alarmed.

ALL actions required by 3-ARP-9-5A, Window 17 were completed.

CRD 34-19 temperature is now 351 °F and stable.

Which one of the following completes both statements?

CRD 34-19 _ (1) _required to be declared SLOW.

Tech Spec LCO 3.1.4, Control Rod Scram Times,_ (2) _met.

[Reference and Illustration Provided]

A. (1) is (2) is Not B. (1) is (2) is C. (1) is Not (2) is Not D. (1) is Not (2) is

59 I slow 55 51 47 I slow 43 I slow 39 I slow 35 slow 31 27 23 I slow 19 slow I slow 15 11 07 I slow 03 I 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58

Q92 Which one of the following completes both statements regarding the Traversing Incore Probe (TIP) system on Unit 2?

In accordance with Tech Spec 3.6.1.3 the TIP ball valves_ (1) _primary containment isolation valves.

_ (2) _ is the procedure that contains guidance for firing a TIP shear valve when a TIP ball valve fails to close.

A. (1) are (2) 2-AOI-64-2E,Traversing Incore Probe Isolation B. (1) are (2) 2-0I-94, Traversing Incore Probe System C. (1) are Not (2) 2-AOI-64-2E,Traversing Incore Probe Isolation D. (1) are Not (2) 2-0I-94, Traversing Incore Probe System

Q93 On 7-1-14, Unit 3 is operating at 100% power.

10:00 Loss ofRPS "A" occurred and the RPS transformer is tagged out.

IfRPS A cannot be restored when is the unit required by Tech Specs to be in MODE 3?

(Assume No Operator actions)

[Reference Provided]

A. 7-1@ 20:00 B. 7-1 @ 23:00 C. 7-2@ 10:00 D. 7-2 @ 22:00

Q94 Unit 1 is MODE 2 with a startup in progress in accordance with 1-GOI-100-1A, Unit Startup.

Reactor Pressure is 955 psig and the first bypass valve is 10% open. The US is preparing to brief the crew on swapping auxiliary steam loads to Main Steam.

Chemistry reports the following reactor water chemistry parameters to the Control Room:

Chlorides: 0.09 ppm

Conductivity: 1.5 ~-tmhos/cm

pH: 5.0 Which ONE ofthe following identifies the minimum required action(s) in accordance with TRM 3 .4.1, Coolant Chemistry Limits?

[Reference Provided]

A. Condition A.1 only B. Condition B.1 only C. Condition C. I only D. Condition D.1

Q95 Which one ofthe following completes both statements as they pertain to EECW/RHRSW in accordance with OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking?

1. When three operable EECW pumps exist prior to a DG becoming inoperable, Then_ (1) _redundant EECW Pump needs to be declared inoperable.

2. There are _ (2) _redundant RHRSW subsystems per unit.

A. (1) one (2) two B. (1) one (2) four C. (1) no (2) two D. (1) no (2) four

Q96 Which ONE ofthe following completes both statements regarding emergency (priority 1) work orders?

In accordance with NPG-SPP-07 .1.4, Work Control Prioritization - On Line, emergency (priority 1) work orders require the approval of the _ (1) _.

In accordance with NPG-SPP-06.1, Work Order Process, planningfor emergency (priority 1) work orders_ (2) _ required prior to work performance.

A. (1) Shift Manager (2) is B. (1) Shift Manager (2) is Not C. (1) Plant Manager (2) is D. (1) Plant Manager (2) is Not

Q97 To comply with Technical Specifications:

Which of the following completes the statements below?

LCO _ (1) _ establishes the allowance to restore inoperable equipment to service to demonstrate its operability.

This allowance _ (2) _ applicable to restoring equipment to service to demonstrate the operability of OTHER equipment.

A. (1) 3.0.5 (2) is also B. (1) 3.0.5 (2) is Not

c. (1) 3.0.4 (2) is also D. (1) 3.0.4 (2) is Not

Q98 An event involving fuel damage has occurred on Unit 1.

The following conditions exist at 09:00:

Stack Noble Gas WRGERM: 7.1 X 109 J.LCi/sec

O-SI-4.8.B.l.a.1, Airborne Effluent Release Rate, Release Fraction: 15

Four areas in the Reactor Building exceed their Max Safe Radiation levels

Site Boundary Radiation Readings not obtained yet, but will be available at 09:30 What is the highest REQUIRED emergency classification at 09:15?

[Reference Provided]

A. Unusual Event per 4.1-U B. Alert per 4.1-A C. Site Area Emergency per 4.1-S D. General Emergency per 4.1-G

Q99 What procedure provides the guidance for abnormal/emergency annunciator response and what are the requirements for returning to normal annunciator response?

A. BFN-ODM-4.20, Strategies for Successful Transient Mitigation; Normal annunciator response will be resumed when abnormal/emergency procedures are exited only.

B. BFN-ODM-4.20, Strategies for Successful Transient Mitigation; Normal annunciator response will be resumed when abnormal/emergency procedures are exited or at SM/US discretion.

C. OPDP-1, Conduct of Operations; Normal annunciator response will be resumed when abnormal/emergency procedures are exited only.

D. OPDP-1, Conduct of Operations; Normal annunciator response will be resumed when abnormal/emergency procedures are exited or at SM/US discretion.

Q 100 Which one ofthe following completes both statements In accordance with EPIP-1 Emergency Classification Procedure?

IF an Emergency Action Level (EAL) for a higher classification was exceeded, but the present situation indicates a lower classification, THEN the higher classification _ (1) _ be declared.

IF an Emergency Action Level (EAL) was exceeded but has now been totally resolved, THEN the NRC _ (2) _ required to be notified.

A. (1) should Not (2) is B. (1) should Not (2) is Not C. (1) should still (2) is D. (1) should still (2) is Not

RO EXAM REFERENCES ROQ9. 2-AOI-74-1, Loss of Shutdown Cooling, Rev 39

(§Eo EXAM REFERENC@)

SRO Q 76. NPG-SPP-03.5, Regulatory Reporting Requirements, Rev. 0010 SRO Q 77. EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE, Rev 50 SRO Q 78. EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE, Rev 50 SRO Q 82. EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE, Rev 50 SRO Q 98. EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE, Rev 50 SRO Q 79. EOI-CURVE 3, Heat Capacity Temp Limit.

SRO Q 84. EOI-CAUTION 1, EOI-CURVE 8 and Table 6.

SRO Q 87. T.S. 3.3.1.2 Source Range Monitor (SRM) Instrumentation SRO Q 88. T.S. 3.3.1.1 Reactor Protection System (RPS) Instrumentation SRO Q 90. T.S. 3.3.6.1 Primary Containment Isolation Instrumentation SRO Q 91. T.S. 3.1.4 Control Rod Scram Times I Core Matrix showing location of slow rods SRO Q 93. T.S. 3.4.5 RCS Leakage Detection Instrumentation SRO Q 94. TR 3.4.1 Coolant Chemistry

NPG-SPP-03.5 Regulatory Reporting Requirements Rev.0010 Page 1 of 100 Quality Related 0 Yes 0No NPG Standard Programs and Processes Validation Date 07-28-2011 Review Frequency 3 years Validated By Henry Lee Effective Date 03-19-2014 Responsible Peer Team/Working Group: Licensing Ed Schrull 03-18-2014 Approved by:

Corporate Functional Area Manager Date

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 6 of 100 Table of Contents 1.0 PURPOSE ................................................................................................................................. 8 2.0 SCOPE ...................................................................................................................................... 8 3.0 PROCESS ................................................................................................................................. 8 3.1 Roles and Responsibilities ......................................................................................................... 8 3.1.1 Corporate Licensing ................................................................................................... 8 3.1.2 Site Licensing ............................................................................................................. 8 3.2 Instructions ................................................................................................................................. 9 3.2.1 Periodic Reports ......................................................................................................... 9 3.2.2 Event or Condition Reporting ..................................................................................... 9 3.2.3 Processing Reports .................................................................................................. 12 4.0 RECORDS ............................................................................................................................... 13 4.1 QA Records ............................................................................................................................. 13 4.2 Non-QA Records ...................................................................................................................... 13 5.0 DEFINITIONS .......................................................................................................................... 13

6.0 REFERENCES

........................................................................................................................ 18 6.1 Source Documents .................................................................................................................. 18 6.1.1 Business Requirements ............................................................................................ 18 6.1.2 Requirements Documents ........................................................................................ 18 6.2 Developmental References ...................................................................................................... 18 Appendix A: Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants ......................................................................................................... 20 Appendix B: Reporting of Events or Conditions Affecting Activities Involving Byproduct, Source or Special Nuclear Material Licenses ................................................................................................................ 35 Appendix C: Reporting of Events or Conditions Affecting Independent Spent Fuel Storage Installation (ISFSI) .............................................................. 41 Appendix D: Site Event Notification Matrix .............................................................................. 48 Appendix E: Other Regulatory Reporting ................................................................................ 51 Appendix F: (Deleted by Revision 20 of SPP-3.5; Reference NGDC PP-13 .......................... 56 Appendix G: Identification, Evaluation and Notifications Under 10 CFR Part 21 ........................................................................................................................... 57

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 7 of 100 Table of Contents {continued)

Appendix H: Reporting of Decommissioning Funding ........................................................... 75 Appendix 1: Communication with the NRC Following A Significant Operational Event. ................................................................................................ 80 Appendix J: Internal Notification of Events Requiring Serious Accident Investigations ....................................................................................................... 82 Appendix K: Registration Requirements for Spent Fuel Storage Cask Placed into Service .............................................................................................. 85 Appendix L: Reporting of 24-Hour Fitness for Duty Events Under 10 CFR 26 .................... 86 Appendix M: Receipt of NRC Emergency Notification System Blast Dial ............................. 94 : NPG-SPP-03.5-1 -NRC Form 361, Event Notification Worksheet ............................................................................................................. 95 Source Notes ...................................................................................................... 100

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 19 of 100 6.2 Developmental References (continued)

NPG-SPP-01.6, Nuclear Power Group Corporate Duty Officer NSDP-1, Safeguards Event Reporting Guidelines

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 22 of 100 Appendix A (Page 3 of 15)

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

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

2. §50.72 (a)(1 )(i)-The declaration of any of the Emergency classes specified in the licensee's approved Emergency Plan.

NOTE

3. §50.72(b ).(1 )) -Any deviation from the plant's Technical Specifications authorized pursuant to §50.54(x).

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

§73.71, 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 G. Review the 'Incident Categorization' section in NPG-SPP-12.8.8.]

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

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 23 of 100 Appendix A (Page 4 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 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 following criteria require 4-hour notification:

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

2. §50.72(b)(2)(iv)(A)- 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-planned 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. §50.72(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.

N PG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 24 of 100 Appendix A (Page 5 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.1 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.

1. §50.72(b)(3)(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. §50.72(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. §50.72(b)(3)(iv)(A)- Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) [see list below], 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(b)(3)(iv)(A) apply are:

NOTE Actuation of the RPS when the reactor is critical is also reportable under §50.72(b)(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) ECCS for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 25 of 100 Appendix A (Page 6 of 15}

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

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 26 of 100 Appendix A (Page 7 of 15)

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

6. §50.72(b)(3)(xiii)- Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system).

E. Follow-up Notification (§50.72(c))

With respect to the telephone notifications made under paragraphs (a) and (b) [§50.72 (a) and §50.72 (b), respectively] of this section [§50.72], in addition 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 (iii) 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.0010 Processes Page 27 of 100 Appendix A (Page 8 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.2 Twenty-Four Hour Notification - NRC Any violation of the requirement contained in specific operating license conditions, shall be reported to NRC in accordance with the license condition.

3.3 Two-Day Notification- NRC

§50.9(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

§50.9.

3.4 Sixty-Day Verbal Report

§50.73(a)(2)(iv)(A) 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 3.5]). This CFR 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 3.2.6.,

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

1. Is not considered an LER.

2. Should identify that the report is being made under §50.73(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.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 28 of 100 Appendix A (Page 9 of 15)

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 licensee's 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 NPG-SPP-03.5 Programs and Rev.0010 Processes Page 29 of 100 Appendix A (Page 10 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.5 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.

D. Licensee Event Reports A written report shall be prepared in accordance with §50.73(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 US NRC, Document Control Desk, Washington, DC 20555. NUREG 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 3.4]) required by §50.73.

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. §50.73(a)(2)(i)(A)- The completion of any nuclear plant shutdown required by the plant's Technical Specifications.

2. §50.73(a)(2)(i)(B)- Any operation or condition which was prohibited by the plant's 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 performed, 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 plant's Technical Specifications authorized pursuant to §50.54(x).

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 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)(ii)(B)- 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 3.4]

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 in more than one system or multiple main steam isolation valves (MSIVs).

N PG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 31 of 100 Appendix A

{Page 12 of 15)

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 (PWRs) including: high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems.

d. ECCS for boiling water reactors (BWRs) 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.

f. PWR auxiliary or emergency feedwater system.

g. Containment heat removal and depressurization systems, including containment spray and fan cooler systems.

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

i. 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

[§50.73(a)(2)(vi)].

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 32 of 100 Appendix A (Page 13 of 15)

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

10. §50.73(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 residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

11. §50.73(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. §50.73(a)(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 entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

13. §50.73(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 NPG-SPP-03.5 Programs and Rev.0010 Processes Page 33 of 100 Appendix A (Page 14 of 15)

Reporting of Events or Conditions Affecting Licensed Nuclear Power Plants 3.5 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 degradation [§50.73(a)(2)(ix)(B)].

14. §50.73(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 releases.

15. 10 CFR 73, Appendix G, paragraph I - If a one hour notification is made in Appendix A, section 3.1.8.4 of this procedure, then a written notification to the NRC is required within 60 days.

16. For reporting a defect found installed in the Plant's Safety Related Equipment, Radioactive Wastes System, and Special Nuclear Material within an LER, §Part 21 NRC Reporting of Defects and Noncompliance, see Appendix Gin 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;

N PG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev.0010 Processes Page 34 of 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 24-hour notification to the County or State of Tennessee; (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. SQN TS Appendix B compliance guidance is provided in the flowchart in NPG-SPP-05.5, Environmental Control, Appendix B.

c. WBN TS Appendix B compliance is met through the procedures referenced in NPG-SPP-05.5.

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

(1) Refer to NPG-SPP-05.5, Environmental Control, Section Compliance with the NRC Appendix B to the Facility Operating License, for additional guidance.

(2) If required, SQN or WBN Site Licensing shall make a written report to the NRC in accordance with the NRC Non-routine Report, TS Appendix B, Subsections 5.4.2, within 30 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.

BROWNS FERRY NUCLEAR PLANT Unit 0 Emergency Plan Implementing Procedure EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE Revision 0050 Quality Related Level of Use: Reference Use Effective Date: 9/8/2014 Responsible Organization: Radiological Emergency Preparedness PREPARED BY: Matthew L. Clark APPROVED BY: Steven M. Bono

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 PAGE 1 OF 205 TABLE OF CONTENTS TABLE OF CONTENTS ........................................................................................................................... 1 SECTION I

1.0 INTRODUCTION

................................................................................................................................ 3 1 .1 Purpose .............................................................................................................................................. 3

2.0 REFERENCES

................................................................................................................................... 3 2.1 Industry Documents ............................................................................................................................ 3 2.2 Plant Instructions ................................................................................................................................ 3 3.0 INSTRUCTIONS .................................................................................................................................4 3.1 Generallnstructions ............................................................................................................................4 3.2 BFN EPIP-1 Overview ........................................................................................................................ 5 4.0 QA Records .......................................................................................................................................6 5.0 GLOSSARY of ABBREVIATIONS, ACRONYMS, AND DEFINITIONS ............................................ 7 6.0 EVENT CLASSIFICATION INDEX ................................................................................................... 15 SECTION II EVENT CLASSIFICATION MATRIX ...................................................................................................... 17 1.0 Reactor ............................................................................................................................................. 17 2.0 Primary Containment ........................................................................................................................ 25 3.0 Secondary Containment .................................................................................................................. 33 4.0 Radioactivity Release ....................................................................................................................... 39 5.0 Loss of Power ...................................................................................................................................45 6.0 Hazards ............................................................................................................................................ 51 7.0 Natural Events ..................................................................................................................................69 8.0 Emergency Director Judgment ......................................................................................................... 77 SECTION Ill BASIS .....................................................................................................................................................87 1.0 Reactor ............................................................................................................................................. 87 2.0 Primary Containment ...................................................................................................................... 107 3.0 Secondary Containment ................................................................................................................ 125 4.0 Radioactivity Release ..................................................................................................................... 134 5.0 Loss of Power ................................................................................................................................. 145 6.0 Hazards .......................................................................................................................................... 155 7.0 Natural Events ................................................................................................................................ 180 8.0 Emergency Director Judgment ....................................................................................................... 187

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 PAGE 2 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 17 OF 205 REACTOR 1.0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 18 OF 205 NOTES 1.1-U 1/1 .1-A1 Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.

1.1-S1 Applicable in Mode 5 when the Reactor Head is installed.

1.1-G2 The reactor will remain subcritical under all conditions without boron when:

111 Any 19 control rods are inserted to position 02, with all other control rods fully inserted.

Ill All control rods except one are inserted to or beyond position 00.

Ill Determined by Reactor Engineering.

CURVES/TABLES:

TABI..E 1.1- G2 .**.

.. ** MINIMUM.STEAM COOLING PRESS (MSCP)

NUMBER OF OPEN MSRVs MSCP (PSIG) 6 or More 190 5 230 4 290

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 19 OF 205 Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel c:

Cavity with irradiated fuel assemblies expected to Pool with irradiated fuel assemblies expected to z remain covered by water. remain covered by water. c:

(I) c:

)>

r-m OPERATING CONDITION: OPERATING CONDITION m

ModeS ALL z

-1 Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel Cavity expected to result in irradiated fuel Storage Pool expected to result in irradiated fuel )>

assemblies being uncovered. assemblies being uncovered. r-m OPERATING CONDITION: OPERATING CONDITION: ~

ModeS ALL Reactor water level can NOT be maintained (I) above -162 inches. (TAF) =i m

m m

u G')

m OPERATING CONDITION:

z 0

Mode 1 or 2 or 3 -<

Either of the following exists:

The reactor will remain subcritical without boron G')

under all conditions, and m

);> Less than 4 MSRVs can be opened, or z

);> Reactor pressure can NOT be restored and m

maintained above Suppression Chamber ~

r-pressure by at least 70 psi.

It has NOT been determined that the reactor will m remain subcritical without boron under all s:

m

~

conditions and unable to restore and maintain MSCP in Table 1.1-G2.

m z

OPERATING CONDITION: 0 Mode 1 or 2 or 3 -<

OPERATING CONDITION:

Mode 1 or 2 or 3

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO Unit 0 EVENT CLASSIFICATION MATRIX PAGE 20 OF 205 NOTES 1.2 Subcritical is defined as reactor power below the heating range and not trending upward.

CURVES/TABLES:

CURVE 1.2-G HEAT CAPACITY TEMP LIMIT I I

---+---t--1--'-t--t:---SA.FEWHEN RX PRESS--

I a.

w 1-a.

cr:

a.

=>

(/)

12 13 14 15 16 17 18 19 SUPPR PL LVL (FT)

ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX PRESS

EPIP*1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 21 OF 205 Reactor coolant activity exceeds 26 11Ci/gm dose c:

equivalent 1-131 (Technical Specification Limits) z as determined by chemistry sample. c:

en c:

)>

r-m OPERATING CONDITION m

ALL z

-1 Failure of RPS automatic scram functions to bring Reactor coolant activity exceeds 300 !lCi/gm dose the reactor subcritical equivalent lodine-131 as determined by chemistry AND sample. )>

r-Manual scram or ARI (automatic or manual) was m successful. ~

OPERATING CONDITION:

Mode 1 or 2 or 3 Failure of automatic scram, manual scram, and en ARI to bring the reactor subcritical. =i m

m sm

u Ci) m z

0 Failure of automatic scram, manual scram, and ARI. Reactor power is above 3% Q m

AND z

m Either of the following conditions exists: ~

r-

Suppression Pool temp exceeds HCTL. m Refer to Curve 1.2-G. s:

Reactor water level can NOT be restored m and maintained at or above -180 inches. ~

m z

0 OPERATING CONDITION: -<

Mode 1 or2

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 22 OF 205 NOTES CURVES/TABLES:

CURVE 1.5-S HEAT CAPACITY TEMP LIMIT I I I

-'--+-c-+--SAF WHEN RX PRESs--*

70 PSIG a.

,_w a.

a::

a.

U) 11.5 12 13 14 15 16 17 18 19 SUPPR PL LVL (FT)

ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX PRESS

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 23 OF 205 c:

z c:

OR en c:

)>

Valid OG PRETREATMENT RADIATION HIGH r-alarm, 1, 2, or 3-RA-90-157A. m m

z

-1

)>

r-m

~

OPERATING CONDITION:

Mode 4 or 5 Suppression Pool temperature, level and RPV en pressure can NOT be maintained in the safe area =i of Curve 1.5-S. m m

sm

0 G) m OPERATING CONDITION:

z 0

Mode 1 or 2 or 3 -<

(i) m z

m

~

r-m sm m

~

z 0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 24 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 25 OF 205 PRIMARY CONTAINMENT 2.0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 26 OF 205 NOTES CURVES/TABLES:

TABLE.2.t-A INDICATIONS QF PRIMARY SYSTEM LEAKAGE

' INTO PRIMARY CONTAINMENT Primary Containment Pressure High Alarm Drywell Floor Drain Sump Pump Excessive Operation Drywell CAM Activity Increasing Drywell Temperature High Alarm Chemistry Sample Radionuclide Comparison To Reactor Water CURVE 2.1-S PRESS SUPPR PRESS Oi

u; 8

~

Ul Ul

~

(L Q;

..0 E

ro

.s:::

0 c:

0

u; 1

~

0..

(/)

13 14 15 16 17 18 Suppression Pool Water Level (FT)

BFN EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 UnitO Rev.0050 EVENT CLASSIFICATION MATRIX PAGE 27 OF 205 c

z cCJ) c

>I m

<m z

-1 AND Indication of Primary System leakage into Primary Containment. Refer to Table 2.1-A.

OPERATING CONDITION:

Drywell or Suppression Chamber CJ) hydrogen concentration at or above 4% =i m

AND m sm Drywell or Suppression Chamber ;:;o oxygen concentration at or above 5%. (i) m OPERATING CONDITION:

z 0

Mode 1 or 2 or 3 -<

Drywell or Suppression Chamber hydrogen concentration at or above 6% G')

rn z

AND rn Drywell or Suppression Chamber ~

I oxygen concentration at or above 5%. rn s:

rn

~

rn OPERATING CONDITION: OPERATING CONDITION: z Mode 1 or 2 or 3 0

Mode 1 or 2 or 3 -<

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 28 OF 205 NOTES CURVES/TABLES:

DRvw~~~Liln;~~.~feVEL~

WITH RCS BARRIER INTACT INSIDE; PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT3 RAD MONITOR RIHR RAD MONITOR RIHR RADMONITOR RIHR 1-RE-90-272A 196 2-RE-90-272A 642 3-RE-90-272A 196 1-RE-90-273A 297 2-RE-90-273A 297 3-RE-90-273A 297 TABLE 2~3.;S1/2;3-G2 DRYWELL RADIATION LEVELS WITH RCS.BARRIER NOT INTACT INSII)E PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT3 RAD MONITOR RIHR RADMONITOR RIHR RAD MONITOR RIHR 1-RE-90-272A 2981 2-RE-90-272A 2263 3-RE-90-272A 2981 1-RE-90-273A 2960 2-RE-90-273A 2960 3-RE-90-273A 2960 TABLE.2.3-G1 DRYWEI..L RADIATION LEVELS WITH RCS BARRfER NOTINTACT INSiDE PRIMARY CONTAINMENT UNIT 1 UNIT2 UNIT3 RAD MONITOR R/HR RADMONITOR RIHR RADMONITOR RIHR 1-RE-90-272A 90091 2-RE-90-272A 68405 3-RE-90-272A 90091 1-RE-90-273A 89450 2-RE-90-273A 89450 3-RE-90-273A 89450 TABlE 2.3/2;5-U

.. INDICATIONS OF LOSS OF PRIMARY CONTAINMENT Unexplained Loss Of Containment Pressure Exceeding 1, 2, or 3-SI-4.7.A.2.a Limits Inability To Isolate Any Line Exiting Containment When Isolation Is Required Venting Irrespective Of Offsite Release Rates Per EOis/SAMGs

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 29 OF 205 c:

z c:

(I) c:

)>

m m

z

-1 2.

Drywell radiation levels at or above the values listed in Table 2.3-A/2.3-S2, with the RCS barrier intact inside Primary Containment. )>

m

~

OPERATING CONDITION:

us 2.3-S2 us Drywell radiation levels at or above the values Drywell radiation levels at or above the values listed in Table 2.3-S1/2.3-G2 with the RCS barrier listed in Table 2.3-A/2.3-S2, with the RCS barrier (I)

NOT intact inside Primary Containment. intact inside Primary Containment, =i AND m Either of the following exists: m

Indications of loss of Primary Containment. sm Refer to Table 2.3/2.5-U. :::0

Primary Containment integrity can NOT be Ci) maintained.

m z

0 OPERATING CONDITION: OPERATING CONDITION: -<

Mode 1 or 2 or 3 2.3-G1 us Drywell radiation levels at or above the values Drywell radiation levels at or above the values G')

listed in Table 2.3-G1 with the RCS barrier NOT listed in Table 2.3-S1/2.3-G2 with the RCS barrier m intact inside Primary Containment. NOT intact inside Primary Containment, z AND m Either of the following exists:

Indications of loss of Primary Containment. .-~

Refer to Table 2.3/2.5-U. m

Primary Containment integrity can NOT be s:

m

~

maintained.

m z(")

OPERATING CONDITION: OPERATING CONDITION:

Mode 1 or 2 or 3 Mode 1 or 2 or 3 -<

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 30 OF 205 NOTES CURVES/TABLES:

' .. TAf3LE 2.3/2.5-U

. INDICATIONS OF LOSS OF*PRIMARY CONTAINMENT Unexplained Loss Of Containment Pressure Exceeding 1, 2, or 3-SI-4. 7 .A.2.a Limits Inability To Isolate Any Line Exiting Containment When Isolation Is Required Venting Irrespective Of Offsite Release Rates Per EOis/SAMGs

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 31 OF 205 2.4-U Inability to maintain Primary Containment c:

pressure boundary. Refer to Table 2.3/2.5-U. z OR c:

(/)

c:

Drywell identified leakage exceeds 40 gpm. )>

r-m OPERATING CONDITION: OPERATING CONDITION: <

m Mode 1 or 2 or 3 Mode 1 or 2 or 3 z

-1

)>

r-m

~

OPERATING CONDITION:

(/)

=t m

m s

m

u G) m z

0 G) m z

m

~

r-m s:

m

~

m z

0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 32 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 33 OF 205 SECONDARY CONTAINMENT 3.0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 34 OF 205 NOTES CURVES/TABLES:

. .. . . TABLE3~1 MAXIMUM SAFE OPERATING AREA TEMPERATURE LIMITS APPLICABLE PANEL 9-21 MAX SAFE OPERATING AREA TEMPERATURE ELEMENTS VALUE°F (UNLESS OTHERWISE NOTED) UNIT 1 UNIT2 UNIT3 RHR AIC Pump Room 74-95A 215 150 155 RHR B/D Pump Room 74-958 150 210 215 HPCI Turbine Area 73-55A 275 270 270 CS AIC Pump and RCIC Turbine Area 71-41A 190 190 190 RCIC Steam Supply Area 71-418, 41C, 41D 195 200 250 HPCI Steam Supply Area 73-558, 55C, 55D 245 240 240 RHR AIC Pump Supply Area 74-95H 245 240 240 RHR B/D Pump Supply Area 74-95G 190 240 240 Main Steam Line Leak Detection High (XA-55-3D-24) Panel 9-3 TIS-1-60A 315 315 315 RHR Valve Room 74-95E 175 170 175 RWCU lsol Logic Channel AlB Temp (XA-55-58-32/33) Panel 9-5 175 170 175 High 69-835A, B, C, D Aux lnst Room RWCU Outbd lsol Vlv Area 69-29F 220 220 220 RWCU HxArea 69-29G 220 220 220 RWCU Hx Exh Duct 69-29H 220 220 220 RWCU Recirc Pump A Area 69-29D 215 215 215 RWCU Recirc Pump B Area 69-29E 215 215 215 RHR AIC Hx Room 74-95C 210 195 200 RHR B/D Hx Room 74-95D 210 195 200 FPC HxArea 74-95F 160 155 155 TABLE .3.1.~G/3~2..G INDICATIONS OFPOTENTIAL OR SIGNIFICANT FUEL ClADDING FAILURE WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT UNIT 1 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 DRYWELL RADIATION 1-RE-90-272A I > 196 R/HR 2-RE-90-272A I > 642 RIHR 3-RE-90-272A I > 196 RIHR 1-RE-90-273A I > 297 R/HR 2-RE-90-273A I > 297 RIHR 3-RE-90-273A I > 297 RIHR Reactor Coolant Activity Reactor Coolant Activity Reactor Coolant Activity

300 !!Ci/gm Dose Equivalent ::::300 !!Ci/gm Dose Equivalent ::::300 !!Ci/gm Dose Equivalent Iodine 131 Iodine 131 Iodine 131

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 35 OF 205 SECONDARY CONTAINMENT TEMPERATURE Description I I I I I c

z c

en c

)>

m m

z

-1 I I I I I

.-m

)>

o

-1 3.1-S I I I TABLE I us I en An unisolable Primary System leak is discharging into Secondary Containment =i m

AND m 3!:

m Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1. ;o G')

m OPERATING CONDITION:

z 0

Mode 1 or 2 or 3 -<

3.1-G I I I TABLE I us I An unisolable Primary System leak is discharging into Secondary Containment G) m AND z

m sg Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1 .-

m AND s:

m Any indication of potential or significant fuel cladding failure exists. Refer to Table 3.1-G/3.2-G with RCS Barrier intact inside Primary Containment.

~

m z

0 OPERATING CONDITION -<

Mode 1 or 2 or 3

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 36 OF 205 NOTES CURVES/TABLES:

TABLE3.2

. MAXIMUM SAFE OPERATINGAREARADIATION LIMITS AREA RADMONITOR MAX SAFE VALUE MRIHR UNIT 1 UNIT2 UNIT3 RHR West Room 90-25A 1000 1000 1000 RHR East Room 90-28A 1000 1000 1000 HPCI Room 90-24A 1000 1000 1000 CS/RCIC Room 90-26A 1000 1000 1000 Core Spray Room 90-27A 1000 1000 1000 Suppr Pool Area 90-29A 1000 1000 1000 CRD-HCU West Area 90-20A 1000 1000 1000 CRD-HCU East Area 90-21A 1000 1000 1000 TIP Drive Area 90-23A 1000 1000 1000 North RWCU System Area 90-13A 1000 1000 1000 South RWCU System Area 90-14A 1000 1000 1000 RWCU System Area 90-9A 1000 1000 1000 MG Set Area 90-4A 1000 1000 1000 Fuel Pool Area 90-1A 1000 1000 1000 Service Fir Area 90-2A 1000 1000 1000 New Fuel Storage 90-3A 1000 N/A N/A TABLE 3.1-G/3.2-G INDICATIONS. OF POTENTIAL .OR SIGNIFICANT FUEL CLADDING FAILURE WITH RCS BARRIER INTACT INSIDE PRIMARY CONTAINMENT *.

UNIT 1 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 DRYWELL RADIATION 1-RE-90-272A I > 196 R/HR 2-RE-90-272A I> 642 RIHR 3-RE-90-272A I > 196 RIHR 1-RE-90-273A I > 297 R/HR 2-RE-90-273A I> 297 R/HR 3-RE-90-273A I > 297 RIHR Reactor Coolant Activity Reactor Coolant Activity Reactor Coolant Activity

=::: 300 !J-Ci/gm Dose Equivalent :=::: 300 !J-Ci/gm Dose Equivalent :=::: 300 !J-Ci/gm Dose Equivalent Iodine 131 Iodine 131 Iodine 131

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 37 OF 205 SECONDARY CONTAINMENT RADIATION Description I I I I I c:

z c:

C/)

c:

)>

r-m m

z

-t 3.2-A I I I I I Any of the following high radiation alarms on Panel 9-3:

1, 2, or 3-RA-90-1A, Fuel Pool Floor Alarm

1, 2, or 3-RA-90-250A, Reactor, Turbine, Refuel Exhaust

1, 2, or 3-RA-90-142A, Reactor Refuel Exhaust

1, 2, or 3-RA-90-140A, Refueling Zone Exhaust )>

r-m AND ;;:o

-1 Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred.

OPERATING CONDITION:

ALL 3.2-S I I I TABLEI us I An unisolable Primary System leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2.

OPERATING CONDITION:

Mode 1 or 2 or 3 3.2-G I I I TABLE( us I An unisolable Primary System leak is discharging into Secondary Containment G')

m AND z m

Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2. ~

r-AND m s:

m Any indication of potential or significant fuel cladding failure exists. Refer to Table 3.1-G/3.2-G with RCS Barrier intact inside Primary Containment. ~

m z

OPERATING CONDITION (")

Mode 1 or 2 or 3 -<

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 38 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 39 OF 205 RADIOACTIVITY RELEASES 4.0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 40 OF 205 NOTES 4.1-U Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the following:

1. Actual field measurements exceed the limits in table 4.1-U

2. O-SI4.8.B.1.a.1 release fraction exceeds 2.0 If neither assessment can be conducted within 60 minutes then the declaration must be made on the valid WRGERMS reading.

4.1-A Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the following:

1. Actual field measurements exceed the limits in table 4.1-A

2. O-SI4.8.B.1.a.1 release fraction exceeds 200 If neither assessment can be conducted within 15 minutes then the declaration must be made on the valid WRGERMS reading.

4.1-5 Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the following methods:

1. Actual field measurements exceed the limits in table 4.1-S.

2. Projected or actual dose assessments exceed 100 mrem TEDE or 500 mrem CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

4.1-G Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either ofthe following methods:

1. Actual field measurements exceed the limits in table 4.1-G.

2. Projected or actual dose assessments exceed 1000 mrem TEDE or 5000 mrem CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

CURVES/TABLES*.

Table4.1-U RELEASE LIMITS FOR UNUSUAL EVENT TYPE MONITORING METHOD LIMIT DURATION 7

Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X 10 J..LCi/sec 1 Hour Gaseous Release Rate O-SI4.8.B.1.a.1 Release Fraction 2.0 1 Hour Site Boundary Radiation Reading Field Assessment Team 0.10 MREM/HR Gamma 1 Hour Table4.1-A RELEASE LIMITS FOR ALERT ..

TYPE MONITORING METHOD LIMIT DURATION 9

Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X 10 J..LCi/sec 15 Minutes Gaseous Release Rate O-SI4.8.B.1.a.1 Release Fraction 200 15 Minutes Site Boundary Radiation Reading Field Assessment Team 10 MREM/HR Gamma 15 Minutes Table4.1-5 RELEASE LIMITS FOR.SITE AREA EMERGENCY TYPE MONITORING METHOD LIMIT DURATION 9

Gaseous Release Rate Stack Noble Gas (WRGERMS) 5.9 X 10 J..LCi/sec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 100 MREM/HR Gamma 1 Hour 7 3 Site Boundary lodine-131 Field Assessment Team 3.9 X 10 " J..LCI/cm 1 Hour Table 4.1-G RELEASE LIMITS FOR GENERAL EMERGENCY TYPE MONITORING METHOD LIMIT DURATION 10 Gaseous Release Rate Stack Noble Gas (WRGERMS) 5.9 X 10 !JCi/sec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 1000 MREM/HR Gamma 1 Hour 3

Site Boundary lodine-131 Field Assessment Team 3.9 X 10 -s !JCI/ em 1 Hour

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 41 OF 205 GASEOUS EFFLUENT Description 4.1-U I I NOTE I TABLEI I c:

Gaseous release exceeds ANY limit and duration in Table 4.1-U. z c:

CJ) c:

.-)>

m OPERATING CONDITION:

m ALL z

-t 4.1-A I I NOTE I TABLEI I Gaseous release exceeds ANY limit and duration in Table 4.1-A.

.-)>m

~

OPERATING CONDITION:

ALL 4.1-S I I NOTE I TABLEI I CJ)

EITHER of the following conditions exists: =t m

Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-S. m s:

m

Dose assessment indicates actual or projected dose consequences ::0 above 100 mrem TEDE or 500 mrem thyroid CDE. G) m z

OPERATING CONDITION: 0 ALL -<

4.1-G I I NOTE I TABLEI I G)

EITHER of the following conditions exists: m z

Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-G.

m

Dose assessment indicates actual or projected dose consequences .-~

above 1000 mrem TEDE or 5000 mrem thyroid CD E.

m s:

m

~

m OPERATING CONDITION z ALL 0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 42 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 43 OF 205 c:

Main Steam Line break outside z Primary Containment with isolation. c:

(I)

AND c:

)>

Release duration exceeds or will exceed r-60 minutes. m m

OPERATING CONDITION: OPERATING CONDITION: z

-t Mode 1 or 2 or 3 ALL Liquid release rate exceeds 2000 times ECL as determined by chemistry sample AND )>

r-m Release duration exceeds or will exceed 15 minutes. ~

OPERATING CONDITION:

ALL (I)

Unisolable Main Steam Line break outside =i Primary Containment. m m

sm

0 Ci) m OPERATING CONDITION:

z 0

Mode 1 or 2 or 3 -<

Ci) m z

m

~

r-m sm

~

m z(")

EMERGENCY CLASSIFICATION' PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 44 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 45 OF 205 LOSS OF POWER 5.0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 46 OF 205 NOTES 5.1-U Loss of normal and alternate supply voltage implies inability to restore voltage from any qualified source to normal or alternate feeder for at least one of the unit specific boards within 15 minutes. At least two boards must be energized from Diesel power to meet this classification. If only one board can be energized and that board has only one source of power then refer to EAL 5.1-A 1 or 5.1-A2.

5.1-A1 Only one source of power (Diesel or Offsite) is available to any one of the listed unit specific 4KV Shutdown Boards. No power is available to the three remaining boards.

5.1-A2 Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in operation 5.1-S would apply.

5.1-S Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in Shutdown or Refuel 5.1-A2 would apply.

5.1-G Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only.

CURVES/TABLES:

Table 5.1 UNIT 4KV SHUTDOWN BOARD APPLICABILITY APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT 1 A, B, C, and D UNIT2 A, B, C, and D UNIT3 3A, 3B, 3C, and 3D

BFN EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 UnitO Rev.0050 EVENT CLASSIFICATION MATRIX PAGE 47 OF 205 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 c:

(/)

AND c:

At least two Diesel Generators supplying power to )>

unit specific 4KV shutdown boards listing in .--

Table 5.1. m OPERATING CONDITION: <

m ALL z

-1

)>

AND Only ONE source of power available to the

.--m remaining board. ~

OPERATING CONDITION: OPERATING CONDITION:

Loss of voltage to ALL unit specific 4KV shutdown (/)

boards from Table 5.1 for greater than 15 minutes. =i m

m sm

0 G) m z

0 OPERATING CONDITION: -<

5.1-G us Loss of voltage to ALL unit specific 4KV shutdown G) boards from Table 5.1 m AND z m

.--~

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. s m

~

m z

OPERATING CONDITION:

0 Mode 1 or 2 or 3

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 48 OF 205 NOTES 5.2 250V DC power voltage below 248 volts constitutes a loss of DC power to the affected board. The voltage readings may be obtained at the 250V Shutdown Battery Board (or the 250V Plant Battery Board) that is feeding the affected board.

CURVES/TABLES:

Table 5.2-U

' UNIT 4KV SHUTDOWN BOARD APPLICABILITY APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT 1 A, B, C, AND D UNIT2 A, B, C, AND D UNIT3 3A, 3B, 3C, AND 3D Table 5.2-S CRITICAL DC POWER AND ESSENTIAL SYSTEMS COMBINATION LOSS OF CRITICAL 250V DC POWER POTENTIALLY RESULTS (Unit Specific Unless Otherwise Noted) IN I Control Power for 4KV Unit Boards A, B, and C Loss of Main Condenser AND AND Control Power for 480V Unit Boards A and B Loss of Both EHC Pumps AND AND Power for Panel 9-9 Cabinet 1 Loss of All Reactor Feed Pumps II Power for 250V DC RMOV Board A Loss of HPCI Ill Power for 250V DC RMOV Board C Loss of RCIC IV Power for 250V DC RMOV Boards A, B, and C Less than 4 MSRVs AND AND Control Power for 4KV Shutdown Boards A, B, C, and D Loss of All RHR Pumps (4KV Shutdown Boards 3A, 3B, 3C, and 3D for Unit 3) And Core Spray Pumps

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 49 OF 205 c:

z c:

(I) c:

Unplanned loss of 250V DC control power to unit )>

specific 480V shutdown boards A and B r-for greater than 15 minutes. m m

OPERATING CONDITION: z

-4 Modes 4 or 5

)>

r-m

~

us Loss of 250V DC power to ALL combinations (I)

(1, II, Ill, and IV) of essential systems from Table 5.2-S for greater than 15 minutes. =i m

m sm

o G')

m z

0 OPERATING CONDITION:

Mode 1 or 2 or 3 Ci) m z

m

~

r-m s

m

~

m z

0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 50 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 51 OF 205 HAZARDS 6.0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 52 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 53 OF 205 c:

Valid, unexpected increase of ANY in-plant ARM z reading to 1000 mrem/hr (except TIP room). c:

en c:

)>

m OPERATING CONDITION: <

m ALL z

-1 Valid, unexpected increase of ANY in-plant ARM Control Room radiation levels greater than reading to 1000 mrem/hr (except TIP room). 15 mrem/hr.

)>

AND Personnel required in the affected area(s).

m

~

OPERATING CONDITION: OPERATING CONDITION:

ALL ALL en

=t m

m sm

o G) m z

0 G) m z

m

.-~

m s:

m

~

m z

(")

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 54 OF 205 NOTES CURVES/TABLES:

EPIP*1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 55 OF 205 c:

Turbine failure resulting in casing penetration z OR c:

CJ)

Significant damage to turbine or generator seals c:

during operation. )>

r m

OPERATING CONDITION:

m Mode 1, or 2 z

-t Control Room Abandonment from entry into Turbine failure resulting in visible structural 1, 2, or 3-AOI-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 r

+Intake Structure +Control Bay m OPERATING CONDITION:

~

OPERATING CONDITION: Mode 1 or 2 ALL Control Room Abandonment from entry into 1, 2, or 3-AOI-100-2 or 0-SSI-16 for ANY Unit CJ)

Control Room =i AND m Control of reactor water level, reactor pressure, m and reactor power (for Modes 1, or 2, or 3) or :s:

m decay heat removal (for Modes 4, or 5) per :::u 1, 2, or 3-AOI-1 00-2 or 0-SSI-16 as applicable, can G)

NOT be established within 20 minutes after m evacuation is initiated.

z 0

OPERATING CONDITION: -<

ALL G) m z

m

~

r-m s

m

~

m z

0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO Unit 0 EVENT CLASSIFICATION MATRIX PAGE 56 OF 205 NOTES CURVES/TABLES:

' Table 6:4:.u1

.. APPLICABLE PLANT AREA Reactor Building Refuel Floor 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 BayEievation 593', 606', And 617' Diesel Generator Buildings (All Elevations)

Turbine Building (All Elevations)

Intake Pumping Station (All Elevations)

Radwaste Building (All Elevations)

Cable Tunnel (Intake To Turbine Building)

Standby Gas Treatment BuildinQ T;:ll)le 6A;;A APPLICABLE PLANT AREA Reactor Building Refuel Floor 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)

Intake Pumping Station (All Elevations)

Cable Tunnel (Intake To Turbine Building)

Standby Gas Treatment Building

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 57 OF 205 c:

Confirmed fire in ANY plant area listed in Unanticipated explosion within the protected area z Table 6.4-U1 resulting in visible damage to ANY permanent c:

(/)

AND structure or equipment. c:

r-NOT extinguished within 15 minutes.

m OPERATING CONDITION: OPERATING CONDITION: <

m ALL ALL z

-t 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 r-m permanent structure of safety systems in ANY plant area listed in Table 6.4-A. ~

OPERATING CONDITION:

ALL

(/)

=i m

m sm

u Ci) m z

0 G) m z

m

~

r-m s:

m

~

m z

0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 58 OF 205 NOTES CURVESITABLES:

Table 6.5/6.6 APPLICABLE PLANT AREA Reactor Building Refuel Floor Control Bay Diesel Generator Buildings Turbine Buildinq Intake Pumping Station Radwaste Buildinq Cable Tunnel (Intake To Turbine Building)

Standby Gas Treatment Buildinq

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 59 OF 205 TOXIC GASES Description 6.5-U I I I TABLEI I EITHER of the following conditions exists: c:

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. c:

CJ)

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. r-m OPERATING CONDITION:

m ALL z

-1 6.5-A I I I TABLEI I ALL of the following conditions exist:

Plant personnel report toxic gas within any building or 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). r-m

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 I CJ)

=i m

m sm

0 Ci) m z

0 I I I I I G')

m z

m

~

r-m s:

m

~

m z

0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 60 OF 205 NOTES CURVES/TABLES:

Table 6.5/6.6 APPLICABLE PLANT AREA Reactor Buildinq Refuel Floor Control Bay Diesel Generator Buildings Turbine Buildinq Intake Pumping Station Radwaste Buildinq Cable Tunnel (Intake To Turbine Building)

Standby Gas Treatment Building

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 61 OF 205 FLAMMABLE GASES Description 6.6-U I I I TABLEI J EITHER of the following conditions exists:

c:

Release of flammable gas within the site boundary in concentrations at or above 25% of the Lower z Explosive Limit (LEL) for any three readings obtained in a 10ft. triangular area as indicated by Fire c:

(/)

Protection or Industrial Safety personnel using appropriate monitoring instrumentation. c:

Confirmed report by local, county, or state officials that a large offsite flammable gas release has )>

occurred within one mile of the site with potential to enter the site boundary in concentrations at or .-

above 25% of the Lower Explosive Limit (LEL). m m

OPERATING CONDITION: z ALL -1 6.6-A I I I TABLEI I Release of flammable gases within any building or structure listed in Table 6.5/6.6 in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10ft. triangular area as indicated by Fire Protection or Industrial Safety personnel using appropriate monitoring instrumentation. .-m

)>

~

OPERATING CONDITION:

ALL I I I I I

(/)

=i m

m sm

u G) m z

0 I I I I I G')

m z

m m

s:

m

0 G')

m z

0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 62 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 63 OF 205

1. A SECURITY CONDITION that does NOT c:

involve a HOSTILE ACTION as reported by z the Security Shift Supervisor. c:

(/)

OR c:

2. A credible Browns Ferry threat notification )>

OR r-

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

m z

OPERATING CONDITION: -1 ALL

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

)>

OR r-m

2. A validated notification from NRC of an airliner ~

attack threat within 30 minutes of the site.

OPERATING CONDITION:

L C/)

A HOSTILE ACTION is occurring or has occurred :::j m

within the PROTECTED AREA as reported by the m Security Shift Supervisor :s::

m

0 G)

OPERATING CONDITION: m ALL z 0

Ci)

1. A HOSTILE ACTION has occurred such that m z

plant personnel are unable to operate m equipment required to maintain safety functions. ~

r m

OR s:

m

2. A HOSTILE ACTION has caused failure of ~

m Spent Fuel Cooling Systems and IMMINENT z 0

fuel damage is likely for a freshly off-loaded reactor -<

core in pool.

OPERATING CONDITION:

ALL

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN UnitO Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 64 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 65 OF 205 VEHICLE CRASH Description 6.8-U I I I I I c:

Vehicle crash (for example; aircraft or barge) into plant structures or systems within the protected area z boundary. c:

CJ) c:

)>

r-m OPERATING CONDITION: <

m ALL z

-1 6.8-A I I I I I Vehicle crash (for example; aircraft or barge) into ANY plant vital area.

)>

r-m

0

-1 OPERATING CONDITION:

ALL I I I I I CJ)

=i m

m

s:

m

0 Ci) m z

0 I I I I I G) m z

m

~

r-m sm

a G) m z

0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 66 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 67 OF 205 SPENT FUEL STORAGE Description 6.9-U I I I I I Damage to a loaded cask CONFINEMENT BOUNDARY from ANY of the following: c z

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

c en

Accident (e.g., dropped cask, tipped over cask, explosion, missile damage, fire 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 z

OPERATING CONDITION: -1 ALL I I I I I

.-)>m

~

I I I I I en

=i m

m s:

m

.tJ G')

m z

0 I I I I I (j) m z

m

.-m~

sm

.tJ (j) m z

0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 68 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 69 OF 205 NATURAL EVENTS 7.0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 70 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 71 OF 205 EARTHQUAKE Description 7.1-U I I I I I Valid annunciation in Unit 1 Control Room, Panei1-XA-55-22C, Window 5, c:

START OF STRONG MOTION ACCELEROGRAPH z c:

C/)

AND c:

)>

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

~

OPERATING CONDITION: m ALL z

-1 7.1-A I I I I I Valid annunciation in the Unit 1 Control Room, Panei1-XA-55-22C, Window 6, 1

/ 2 SSE RESPONSE SPECTRUM EXCEEDED AND )>

r-m Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

~

OPERATING CONDITION:

ALL I I I I I C/)

=i m

m sm

u (i) m z

0 I I I I I G) m z

m

~

r-m s:

m

u G) m z

0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO UnitO EVENT CLASSIFICATION MATRIX PAGE 72 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 73 OF 205 TORNADO I HIGH WINDS Description 7.2-U I I I I I c:

Report by plant personnel of tornado striking within the protected area boundary. z c::

C/)

c::

r-m OPERATING CONDITION: <

m ALL z

-t 7.2-A I I I I I Tornado striking plant vital area OR )>

r-Onsite wind speed above 90 MPH as indicated using the meteorological data screen of the m

u Integrated Computer System (ICS). -t OPERATING CONDITION:

ALL I I I I I C/)

=i m

m sm

u G) m z

0 I I I I I G) m z

m

~

r-m s:

m

u G) m z

0

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 74 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 75 OF 205 FLOOD Description 7.3-U I I I I I Wheeler Lake level exceeds or is predicted to exceed elevation 565 feet. c:

z AND c:

(/)

c:

Water entering permanent plant structures due to flooding. )>

r-m OPERATING CONDITION: <

m ALL z

-1 7.3-A I I I I I Wheeler Lake level exceeds or is predicted to exceed elevation 565 feet.

AND

)>

r-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 I

(/)

=i m

m sm

u Ci) m z

0 I I I I I (i) m z

m

~

r-m s

m

u (i) m z

0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 76 OF 205 THIS PAGE INTENTIONALLY BLANK

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Unit 0 Rev.OOSO EVENT CLASSIFICATION MATRIX PAGE 77 OF 205 EMERGENCY DIRECTOR JUDGMENT 8.0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 78 OF 205 NOTES CURVES/TABLES:

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.OOSO Unit 0 EVENT CLASSIFICATION MATRIX PAGE 79 OF 205 TECHNICAL SPECIFICATIONS Description 8.1-U I I I I I c:

Inability to reach required shutdown condition (Mode 3 or Mode 4) within z Technical Specification Limiting Conditions for Operation (LCO) limits. c:

CJ) c:

)>

m OPERATING CONDITION: <

m Mode 1 or 2 or 3 z

-1 I I I I I

.-)>m

a

-1 I I I I I CJ)

=i m

m s:

m

a Ci) m z

0 I I I I I Ci) m z

m

.-~m s:

m

o Ci) m z

0

I t EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 80 OF 205 NOTES CURVES/TABLES:

Table8.2-U LOSS OF COMMUNICATIONS Onsite Communications Offsite Communication Plant Phone System Node 1 Bell Lines Two-Way Radio System Digital Microwave (NSS 1, NSS 2, OPS F2, and OPS F4)

Sound Power Phones NRC Emergency Telecommunication System Nextel Communication System Cellular Phones (If Available)

Health Physics Radio Network

' I EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 UnitO EVENT CLASSIFICATION MATRIX PAGE 81 OF 205 LOSS OF COMMUNICATION Description 8.2-U I I ITABLE I I Unplanned loss of onsite communication listed in Table 8.2-U that defeats the Plant Operations Staff's c:

ability to perform routine operations z c:

C/)

OR c:

r-Unplanned loss of ALL off-site communication listed in Table 8.2-U. m m

OPERATING CONDITOIN:

z

-1 ALL I I I I I r-m

a

-1 I I J J I C/)

=i m

m sm

a G) m z

0 I I I I I G) m z

m

~

r-m sm

a G) m z

0

EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 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.

CURVES/TABLES:

Table8.3-S APPLICABLE SAFETY FUNCTIONS Reactor Power Reactor Pressure Reactor Level Subcriticality Drywell Temperature Drywell Pressure Suppression Chamber Pressure Suppression Pool Temperature Suppression Pool Level

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 83 OF 205 LOSS OF ASSESSMENT CAPABILITY Description 8.3-U I I I I I Unplanned loss of most or all safety system annunciators or indicators which causes a significant c:

loss of plant assessment capability for greater than 15 minutes z AND c:

CJ)

Compensatory non-alarming safety system indications are available (SPDS, ICS) c:

AND )>

In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant. r-m OPERATING CONDITION: <

m MODE 1, or2, or3 z

-1 8.3-A I I NOTE 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 AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant

)>

AND r-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 1, or 2, or 3 8.3-S I I NOTE I TABLEI I Loss of most or all annunciators associated with safety systems CJ)

AND =i Compensatory non-alarming safety system indications are NOT available (SPDS, ICS) m AND m Indications needed to monitor safety functions are NOT available (Refer to Table 8.3-S) sm AND ;:o A significant transient is in progress. G')

m OPERATING CONDITION:

z 0

MODE 1, or 2, or 3 -<

I I I I I Ci) m z

m

~

r-m s:

m

~

m z

0

l I EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 BFN Rev.0050 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.

CURVES/TABLES:

Table8.4~U OTHER

' . EXAMPLE UNUSUALEVENTS .*.

Plant Transient Response 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

' 1 EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 85 OF 205 OTHER Description 8.4-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 z systems occurs. Refer to Table 8.4-U for examples.

c (I) c OR .-

)>

m Any loss or any potential loss of containment. <

m OPERATING CONDITION:

z

-1 ALL 8.4-A I I NOTE I I J 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 .-

m Any loss or potential loss of fuel cladding or RCS pressure boundary. ~

OPERATING CONDITION:

ALL 8.4-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 (I) 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 =i m

result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site m boundary. sm OR Any loss or potential loss of both fuel cladding and RCS pressure boundary. :::0 G')

OR m Potential loss of either fuel cladding or RCS pressure boundary and loss of any additional barrier. z 0

OPERATING CONDITION:

ALL 8.4-G I I NOTE I I I Events are in process or have occurred which involve actual or imminent substantial core degradation or G) melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss m of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective z m

Action Guideline exposure levels offsite for more than the immediate site area.

OR

.-~

m sm Loss of any two barriers and potential loss of third barrier.

~

m z

OPERATING CONDITION: 0 ALL -<

EPIP-1 BFN EMERGENCY CLASSIFICATION PROCEDURE Rev.0050 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 86 OF 205 THIS PAGE INTENTIONALLY BLANK

Curve 3 Heat Capacity Temp Limit 260 250 240 u.. 230 0

a. 220 E

Q) 210 1-

a. 200 I-a.

a.

190

J

(/) 180 170 160 150 11_5 12 13 14 15 16 17 18 19 Suppr PI Lvl (ft}

-ACTION REQUIRED if above curve for existing RPV press

CAUTIONS CAUTION#1

An RPV water lvl instrument may be used to determine or trend lvl only when it reads above the Minimum Indicated Lvl associated with the highest max OW or sc run temp

If OW temps or sc area temps (fable 6), as applicable, are outside the safe region of Curve 8, the associated instrument may be unreliable due to boiling in the run MINIMUM MAX OW RUN TEMP MAXSC INSTRUMENT RANGE INDICATED (FROM XR-64-50 RUN TEMP LVL OR TI-64-52AB) (FROM TABLE 6) on scale NJA below150

-145 NJA 151 to 200 Emergency LI-3-58A/B -140 NJA 201 to250

-155 to +60

-130 NJA 251 to 300

-120 NJA 301 to 350 Ll-3-53 on scale NJA beiOW150 Ll-3-60 Normal +5 NJA 151 to 200 Ll-3-206 +15 NJA 201 to250 Oto +60 Ll-3-253 +20 NJA 251 to 300 LI-3-208A, B, C, 0 +30 NJA 301 to 350 Post Ll-3-52 Accident on scale NJA N/A LI-3-62A -268to +32

+10 Below 100 N/A

+15 100 to 150 NJA Shutdown +20 151 to 200 NJA Ll-3-55 Floodup +30 201 to 250 NJA Oto +500 +40 251 to 300 NJA

+50 301 to 350 NJA

+65 351 to400 N/A

Curve 8 r- RPV Saturation Temp 1

I 400 I 380 I

I ~ 360

<Ill I § 340 0::

I 1: 320 I (!)

I §..... 300 I ~ 280 I (ij 260 I (!)

I ~ 240 E

-1 ~ 220 I

200~~~~~~+=~~~~

I 0 50 100 150 200 250 I

RPV Press (psig) *Constant above 250 psig I

I I

I Table 6 Secondary Cntmt Instrument Runs INSTRUMENT SC TEMP ELEMENTS AND LOCATIONS El621 6593 El565 RWCUHXRM (74-95F) {14-95C and D) (69-835A thru D) (69-29F G H)

LI-3-58A "F Of N/A OF Ll-3-586 "F Of N/A N/A Ll-3-53 "F Of N/A OF Ll-3-60 "F Of N/A NIA Ll-3-206 "F Of N/A Of Ll-3-253 "F Of N/A NIA Ll-3-52 "F Of "F NIA LI-3-62A "F Of "F NIA Ll-3-55 "F Of N/A N/A LI-3-208A, B "'F Of N/A Of LI-3-208C, D "F Of N/A NIA

SRM Instrumentation 3.3.1.2 3.3 INSTRUMENTATION 3.3.1.2 Source Range Monitor (SRM) Instrumentation LCO 3.3.1.2 The SRM instrumentation in Table 3.3.1.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.2-1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required SRMs to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SRMs inoperable in OPERABLE status.

MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.

B. Three required SRMs B.1 Suspend control rod Immediately inoperable in MODE 2 withdrawal.

with IRMs on Range 2 or below.

C. Required Action and C.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 or B not met.

(continued)

BFN-UNIT 1 3.3-9 Amendment No. 234

SRM Instrumentation 3.3.1.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more required D.1 Fully insert all insertable 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SRMs inoperable in control rods.

MODE 3 or4.

AND D.2 Place reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the shutdown position.

E. One or more required E.1 Suspend CORE Immediately SRMs inoperable in ALTERATIONS except for MODE 5. control rod insertion.

AND E.2 Initiate action to fully Immediately insert all insertable control rods in core cells containing one or more fuel assemblies.

BFN-UNIT 1 3.3-10 Amendment No. 234

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS

N()TE---------------------------------------------------

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable M()DE or other specified conditions.

SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2.2 -------------------------N()TES------------------------

1. ()nly required to be met during C()RE ALTERATI()NS.

2. ()ne SRM may be used to satisfy more than one of the following.

Verify an ()PERABLE SRM detector is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months located in:

a. The fueled region;

b. The core quadrant where C()RE ALTERATI()NS are being performed, when the associated SRM is included in the fueled region; and

c. A core quadrant adjacent to where C()RE ALTERATI()NS are being performed, when the associated SRM is included in the fueled region.

SR 3.3.1.2.3 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (continued)

BFN-UNIT 1 3.3-11 Amendment No. 234

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2.4 --------------------------N()TE-------------------------

Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.

Verify count rate is ;::: 3.0 cps with a signal to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during noise ratio;::: 3:1. C()RE ALTERATI()NS 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.1.2.5 Perform CHANNEL FUNCTI()NAL TEST and 7 days determination of signal to noise ratio.

SR 3.3.1.2.6 --------------------------N()TE-------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRMs on Range 2 or below.

Perform CHANNEL FUNCTI()NAL TEST and 31 days determination of signal to noise ratio.

(continued)

BFN-UNIT 1 3.3-12 Amendment No. 234

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.3.1.2.7 -------------------------N()TES------------------------

1. Neutron detectors are excluded.

2. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRMs on Range 2 or below.

Perform CHANNEL CALIBRATI()N. 92 days BFN-UNIT 1 3.3-13 Amendment No. 234

SRM Instrumentation 3.3.1.2 Table 3.3.1.2-1 (page 1 of 1)

Source Range Monitor Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTION SPECIFIED CONDITIONS CHANNELS REQUIREMENTS

1. Source Range Monitor 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 3,4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 5 2(b)(c) SR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.7 (a) With IRMs on Range 2 or below.

(b) Only one SRM channel is required to be OPERABLE during spiral offload or reload when the fueled region includes only that SRM detector.

(c) Special movable detectors may be used in place of SRMs if connected to normal SRM circuits.

BFN-UNIT 1 3.3-14 Amendment No. 234

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 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 -------------N0 T E-------------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

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

B. -------------NOTE------------ B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for system in trip.

Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR One or more Functions B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with one or more required trip.

channels inoperable in both trip systems.

(continued)

BFN-UNIT 1 3.3-1 Amendment No. 234, 262, 266 December 29, 2006

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

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

not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, or Table 3.3.1.1-1 for the C not met. channel.

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

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully Immediately Action D.1 and insert all insertable referenced in control rods in core cells Table 3.3.1.1-1. containing one or more fuel assemblies.

BFN-UNIT 1 3.3-2 Amendment No. 2-34,--262 September 27, 2006

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required 1.1 Initiate alternate method 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and to detect and suppress referenced in thermal hydraulic Table 3.3.1.1-1. instability oscillations.

J. Required Action and J.1 Be in Mode 2. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time of Condition I not met.

BFN-UNIT 1 3.3-2a Amendment No.266 December 29, 2006

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

N()TES--------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.1.1.2 --------------------------N()TE-------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL P()WER ~ 25% RTP.

Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is

~ 2% RTP while operating at ~ 25% RTP.

SR 3.3.1.1.3 --------------------------N()TE-------------------------

Not required to be performed when entering M()DE 2 from M()DE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering M()DE 2.

Perform CHANNEL FUNCTI()NAL TEST. 7 days (continued)

BFN-UNIT 1 3.3-3 Amendment No. 2d4,--262 September 27, 2006

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.6 --------------------------NOTE-------------------------

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

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

1. Neutron detectors are excluded.

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

Perform CHANNEL CALIBRATION. 92 days (continued)

BFN-UNIT 1 3.3-4 Amendment No. ~262 September 27, 2006

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 0 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.11 (Deleted)

SR 3.3.1 .1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 --------------------------NOTE-------------------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST.

SR 3.3.1.1.15 Verify Turbine Stop Valve- Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure- Low Functions are not bypassed when THERMAL POWER is 2 30% RTP.

SR 3.3.1.1.16 --------------------------NOTE-------------------------

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

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is 2 25% and recirculation drive flow is< 60% of rated recirculation drive flow.

BFN-UNIT 1 3.3-5 Amendment No. 234, 262, 263, 266 December 29, 2006

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. Average Power Range Monitors

a. Neutron Flux - High, 2 3(b) G SR 3.3.1.1.1 I Setdown SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16

b. Flow Biased Simulated 3(b) F SR 3.3.1.1.1 Thermal Power - High SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16

c. Neutron Flux - High 3(b) F SR 3.3.1.1.1 I SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM channel provides inputs to both trip systems.

BFN-UNIT 1 3.3-6 Amendment No. 236, 262, 269 March 06, 2007

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. Average Power Range Monitors (continued)

d. lnop 1,2 3(b) G SR 3.3.1.1.16 NA

e. 2-0ut-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.16

f. OPRM Upscale 3(b) SR 3.3.1.1.1 NA SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17

* *

I

=

* *

I BFN-UNIT 1 3.3-7 Amendment No. 269 234,262,259,257,258,266 March 06, 2007

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1 I

I I

I BFN-UNIT 1 3.3-8 Amendment No. ~257 September 14, 2006

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

APPLICABILITY: According to Table 3.3.6.1-1.

BFN-UNIT 2 3.3-53 Amendment No. 253

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS

NOTE:---------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION RE:QUIRE:D ACTION COMPLE:TION TIME:

A. One or more required A.1 --------------N0 TE: ------------

channels inoperable. Only applicable for Function 1.d if two or more channels are inoperable.

Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Functions 2.a, 2.b, 5.h, 6.b, and 6.c AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for Functions other than Functions 2.a, 2.b, 5.h, 6.b, and 6.c AND A.2 --------------N0 T E: ------------

Only applicable for Function 1.d when 15 of 16 channels are OPE:RABLE:.

Place channel in trip. 30 days (continued)

BFN-UNIT 2 3.3-54 Amendment No. 253

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

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

not maintained. OR 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Function 1.d when normal ventilation is not available C. Required Action and C.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A or B Table 3.3.6.1-1 for the not met. channel.

D. As required by Required D.1 Isolate associated Main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action C.1 and Steam Line (MSL).

referenced in Table 3.3.6.1-1.

OR D.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND D.2.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

BFN-UNIT 2 3.3-55 Amendment No. 253

Primary Containment Isolation Instrumentation 3.3.6.1 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action C.1 and referenced in Table 3.3.6.1-1.

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

referenced in Table 3.3.6.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action C.1 and referenced in AND Table 3.3.6.1-1.

G.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Required Action and associated Completion Time for Condition F not met.

(continued)

BFN-UNIT 2 3.3-56 Amendment No. 253

Primary Containment Isolation Instrumentation 3.3.6.1 A CTIONS (continued)

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

Table 3.3.6.1-1.

OR H.2 Isolate the Reactor Water 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Cleanup System.

I. As required by Required 1.1 Initiate action to restore Immediately Action C.1 and channel to OPERABLE referenced in status.

Table 3.3.6.1-1.

OR 1.2 Initiate action to isolate Immediately the Residual Heat Removal (RHR)

Shutdown Cooling System.

BFN-UNIT 2 3.3-57 Amendment No. 253

Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS

N()TES--------------------------------------------------

1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.6.1.2 Perform CHANNEL FUNCTI()NAL TEST. 92 days SR 3.3.6.1 .3 Perform CHANNEL CALIBRATI()N. 92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATI()N. 122 days SR 3.3.6.1.5 Perform CHANNEL CALIBRATI()N. 24 months SR 3.3.6.1.6 Perform L()GIC SYSTEM FUNCTI()NAL 24 months TEST.

BFN-UNIT 2 3.3-58 Amendment No. 255 November 30, 1998

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.1

1. Main Steam Line Isolation

a. Reactor Vessel Water 1,2,3 2 D SR 3.3.6.1.1 Level - Low Low Low, SR 3.3.6.1.2 Level 1 SR 3.3.6.1.5 SR 3.3.6.1.6

b. Main Steam Line Pressure 2 E SR 3.3.6.1.2 I

- Low(c) SR 3.3.6.1.5 SR 3.3.6.1.6

c. Main Steam Line Flow - 1,2,3 2 per D SR 3.3.6.1.1 High MSL SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6

d. Main Steam Tunnel 1,2,3 8 D SR 3.3.6.1.2 I Temperature - High SR 3.3.6.1.5 SR 3.3.6.1.6

2. Primary Containment Isolation

a. Reactor Vessel Water 1,2,3 2 G SR 3.3.6.1.1 Level - Low, Level3 SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6

b. Drywell Pressure - High 1,2,3 2 G SR 3.3.6.1.2 I SR 3.3.6.1.5 SR 3.3.6.1.6

3. High Pressure Coolant Injection (HPCI) System Isolation

a. HPCI Steam Line Flow - 1,2,3 F SR 3.3.6.1.2 I High SR 3.3.6.1.5 SR 3.3.6.1.6

b. HPCI Steam Supply Line 1,2,3 3 F SR 3.3.6.1.2 I Pressure - Low SR 3.3.6.1.5 SR 3.3.6.1.6

c. HPCI Turbine 1,2,3 3 F SR 3.3.6.1.2 I Exhaust Diaphragm SR 3.3.6.1.5 Pressure - High SR 3.3.6.1.6 (c) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

BFN-UNIT 2 3.3-59 Amendment No. 253, 260, 296 September 14, 2006

tt Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.1

3. HPCI System Isolation (continued)

d. HPCI Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 HPCI Pump Room Area SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

e. HPCI Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 I Torus Area (Exit) SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

f. HPCI Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 I Torus Area (Midway)

SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

g. HPCI Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 Torus Area (Entry) SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line Flow - 1,2,3 F SR 3.3.6.1.2 I High SR 3.3.6.1.5 SR 3.3.6.1.6

b. RCIC Steam Supply Line 1,2,3 3 F SR 3.3.6.1.2 I Pressure - Low SR 3.3.6.1.5 SR 3.3.6.1.6

c. RCIC Turbine 1,2,3 3 F SR 3.3.6.1.2 I Exhaust Diaphragm SR 3.3.6.1.5 Pressure - High SR 3.3.6.1.6

d. RCIC Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 I RCIC Pump Room Area SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

e. RCIC Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 Torus Area (Exit) SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

f. RCIC Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 I Torus Area (Midway) SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6

g. RCIC Steam Line Space 1,2,3 2 F SR 3.3.6.1.2 I Torus Area (Entry) SR 3.3.6.1.5 Temperature - High SR 3.3.6.1.6 (continued)

BFN-UNIT 2 3.3-60 Amendment No. ~297 September 21, 2006

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 3)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION C.1

5. Reactor Water Cleanup (RWCU) System Isolation

a. Main Steam Valve Vault 1,2,3 2 F SR 3.3.6.1.2 I Area Temperature- High SR 3.3.6.1.5 SR 3.3.6.1.6

b. Pipe Trench Area 1,2,3 2 F SR 3.3.6.1.2 I Temperature- High SR 3.3.6.1.5 SR 3.3.6.1.6

c. Pump Room A Area 1,2,3 2 F SR 3.3.6.1.2 I Temperature- High SR 3.3.6.1.5 SR 3.3.6.1.6

d. Pump Room B Area Temperature- High 1,2,3 2 F SR 3.3.6.1.2 I SR 3.3.6.1.5 SR 3.3.6.1.6

e. Heat Exchanger Room 1,2,3 2 F SR 3.3.6.1.2 I Area (West Wall) SR 3.3.6.1.5 Temperature- High SR 3.3.6.1.6

f. Heat Exchanger Room 1,2,3 2 F SR 3.3.6.1.2 I Area (East Wall) SR 3.3.6.1.5 Temperature- High SR 3.3.6.1.6

g. SLC System Initiation 1,2 1(a) H SR 3.3.6.1.6 NA

h. Reactor Vessel Water 1,2,3 2 F SR 3.3.6.1.1 Level - Low, Level3 SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6

6. Shutdown Cooling System Isolation

a. Reactor Steam Dome 1,2,3 F SR 3.3.6.1.2 I Pressure - High SR 3.3.6.1.5 SR 3.3.6.1.6

b. Reactor Vessel Water 3,4,5 2(b) SR 3.3.6.1.1 Level- Low, Level3 SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6

c. Drywell Pressure- High 1,2,3 2 F SR 3.3.6.1.2 I SR 3.3.6.1.5 SR 3.3.6.1.6 (b) Only one channel per trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained.

BFN-UNIT 2 3.3-61 Amendment No. 297 Amendment No. 253, 254, 260, 277, September 21, 2006

Control Rod Scram Times 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 13 OPERABLE control rods shall be "slow," in accordance with Table 3.1.4-1; and

b. No more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months not met.

BFN-UNIT 3 3.1-12 Amendment No. 212

Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS

N()TE-----------------------------------------------------

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from .the associated scram accumulator.

SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is within Prior to the limits of Table 3.1.4-1 with reactor steam exceeding dome pressure ~ 800 psig. 40% RTP after each reactor shutdown

~ 120 days SR 3.1.4.2 Verify, for a representative sample, each 200 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure ~ 800 psig. M()DE1 (continued)

BFN-UNIT 3 3.1-13 Amendment No. 212, 226, 253 January 09, 2006

Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.4.3 Verify for each affected control rod scram Prior to declaring time is within the limits of Table 3.1.4-1 with control rod any reactor steam dome pressure. OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time is Prior to within the limits of Table 3.1.4-1 with reactor exceeding steam dome pressure :2: 800 psig. 40% RTP after fuel movement within the affected core cell Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time BFN-UNIT 3 3.1-14 Amendment No. ~ 226 November 21, 2000

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)

Control Rod Scram Times

~()TE:S----------------------------------------------------

1. ()PE:RABLE: control rods with scram times not within the limits of this Table are considered "slow."

2. E:nter applicable Conditions and Required Actions of LC() 3.1.3, "Control Rod

()PE:RABILITY," for control rods with scram times> 7 seconds to notch position 06.

These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow."

SCRAM TIME:s(a)(b)

(seconds)

N()TCH P()SITI()~ RE:ACT()R STE:AM D()ME: PRE:SSURE:

~ 800 psig 46 0.45 36 1.08 26 1.84 06 3.36 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

(b) Scram times as a function of reactor steam dome pressure, when< 800 psig are within established limits.

BFN-UNIT 3 3.1-15 Amendment No. 212

59 slow 55 51 47 slow 43 I slow 39 1 slow 35 slow I 31 27 23 I slow 19 slow I slow 15 11 07 slow 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58

RCS Leakage Detection Instrumentation 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Leakage Detection Instrumentation LCO 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:

a. Drywell floor drain sump monitoring system; and

b. One channel of either primary containment atmospheric particulate or atmospheric gaseous monitoring system.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell floor drain sump A.1 Restore drywell floor 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months monitoring system drain sump monitoring inoperable. system to OPERABLE status.

(continued)

BFN-UNIT 3 3.4-12 Amendment No. 212

RCS Leakage Detection Instrumentation 3.4.5 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required primary B.1 Analyze grab samples of Once per containment atmospheric primary containment 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring system atmosphere.

inoperable.

AND B.2 Restore required primary 30 days containment atmospheric monitoring system to OPERABLE status.

C. Required Action and C.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 or B AND not met.

C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. All required leakage D.1 Enter LCO 3.0.3. Immediately detection systems inoperable.

BFN-UNIT 3 3.4-13 Amendment No. ~ 244 December 1, 2003

RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months primary containment atmospheric monitoring system instrumentation.

SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of 31 days required primary containment atmospheric monitoring system instrumentation.

SR 3.4.5.3 Perform a CHANNEL CALIBRATION of 184 days required drywell sump flow integrator instrumentation.

SR 3.4.5.4 Perform a CHANNEL CALIBRATION of 24 months required leakage detection system instrumentation.

BFN-UNIT 3 3.4-14 Amendment No. 215 November 30, 1998

Coolant Chemistry TR 3.4.1 TR 3.4 REACTOR COOLANT SYSTEM TR 3.4.1 Coolant Chemistry LCO 3.4.1 Reactor coolant chemistry shall be maintained within the limits of Table 3.4.1-1.

APPLICABILITY: According to Table 3.4.1-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Conductivity greater A.1 Verify by administrative Immediately than the limit of means that conductivity Table 3.4.1-1 has not been > 1.0 Column B but ~ 10 1-1mho/cm at 25°C for > 2 1-1mho/cm at 25°C. weeks in the past year.

B. Chloride concentration B.1 Verify by administrative Immediately greater than the limit means that chloride of Table 3.4.1-1 concentration has not Column BorE but been > 0.2 ppm for > 2

~ 0.5 ppm. weeks in the past year.

C. pH not within limits of C.1 Restore pH to within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Table 3.4.1-1 limits.

Column A, B, and E.

(continued)

BFN-UNIT 3 3.4-1 TRM Revision G., 21 March 13, 2001

Coolant Chemistry TR 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate an orderly Immediately associated Completion shutdown.

Time of Conditions A, B, or C not met. AND OR D.2 Be in MODE 4. As rapidly as cooldown rate Conductivity> 10 permits

!lmho/cm at 25°C.

OR Chloride concentration

> 0.5 ppm.

OR Conductivity or chloride concentration limits of Table 3.4.1-1 Column A exceeded.

E. Coolant chemistry E.1 Initiate action to restore Immediately limits of Table 3.4.1-1 coolant chemistry within Column C, D, or E limits.

exceeded.

BFN-UNIT 3 3.4-2 TRM Revisions 0, 16, 19 and 32 August15,2002

Coolant Chemistry TR 3.4.1

~()TE:-------------------------------------------------------

When there is no fuel in the reactor vessel, sampling of reactor coolant chemistry at Technical Requirement frequency is not required.

TE:CH~ICAL SURVE:ILLA~CE: RE:QUIRE:ME:~TS SURVE:ILLA~CE: FRE:QUE:~CY TSR 3.4.1.1 ------------------------- ~ () TE: ------------------------- Continuously

~ot required when there is no fuel in the reactor vessel. ()R 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when Monitor reactor coolant conductivity. the continuous conductivity monitor is inoperable and the reactor is not in M()DE: 4 or 5 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the continuous conductivity monitor is inoperable and the reactor is in M()DE: 4 or 5 TSR 3.4.1.2 ------------------------- ~ () TE: ------------------------- 7 days

~ot required when there is no fuel in the reactor vessel. A~D 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Check the continuous conductivity monitor whenever the with an in-line flow cell. reactor coolant conductivity is

>1.0 ~-tmho/cm at 25°C (continued)

BF~-U~IT 3 3.4-3 TRM Revision 0

Coolant Chemistry TR 3.4.1 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.4.1.3 Verify reactor coolant conductivity and Once during chloride concentration within limits of startup prior to Table 3.4.1-1 Column A. pressurizing the reactor above atmospheric pressure TSR 3.4.1.4 -------------------------N0 TE ------------------------- 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months Only required when the reactor is operating in MODES 1 or 2. AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Verify chloride ion content and pH within the whenever the limits of Table 3.4.1-1. reactor conductivity is

>1.0 !lmho/cm at 25°C (not required for Column E.)

TSR 3.4.1.5 -------------------------NOTE-------------------------

Only required when the reactor is not pressurized with fuel in the reactor vessel.

Verify conductivity, chloride ion content, and 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months pH within the limits of Table 3.4.1-1.

BFN-UNIT 3 3.4-4 TRM Revision G, 21 March 13, 2001

Coolant Chemistry TR 3.4.1 Table 3.4.1-1 Coolant Chemistry Limits( 1)

COLUMNA COLUMN B COLUMN C 2 3 COLUMN 0( ) COLUMN E( )

APPLICABLE APPLICABLE CONDITION APPLICABLE CONDITION APPLICABLE CONDITION APPLICABLE CONDITION CHEMISTRY CONDITION Steaming Rates Reactor Not Pressurized With Fuel In Noble Metal Chemical Application and Operation of HWC Following Noble PARAMETERS Prior To Startup > 100,000 lb/hr Reactor Vessel, Except During Subsequent Reactor Coolant Cleanup Metal Chemical Application And At Steaming Startup Condition Rates

< 100,000 lb/hr CHLORIDE :-: ; 0.1 :-: ; 0.2 :-: ; 0.5 :-: ; 0.1 :<>0.2 (ppm)

CONDUCTIVITY :<>2.0 :<>1.0 :<>10.0 :<>20.0 :<>2.0 (J.lmho/cm at 25°C}

pH 5.6-8.6 5.6-8.6 5.3-8.6 4.3-9.9 5.6-8.8 (1)

When there is no fuel in the reactor vessel, Technical Requirement reactor coolant chemistry limits do not apply.

(2)

During the Noble Metal Chemical Application and subsequent reactor coolant cleanup, CONDITIONS A, B, C, and D (including Required Actions and Completion Times) do not apply.

(3)

During operation of HWC following the Noble Metal Chemical Application, CONDITION A (including Required Action and Completion Time) does not apply.

BFN-UNIT 3 3.4-5 TRM Revision 0-;-4e, 21 March 13, 2001

ML15040A581

Text

6.0 REFERENCES

1.0 INTRODUCTION

2.0 REFERENCES

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