ML020460320

ERP-101, Revision 23, Classification of Emergencies, and ERP-101 Bases, Revision 3, PBAPS EAL Technical Basis Manual.

ML020460320
Person / Time
Site:	Peach Bottom
Issue date:	02/05/2002
From:	Gallagher M Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation, Office of Nuclear Security and Incident Response
References
Download: ML020460320 (172)

Text

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Exelon Nuclear www.exeloncorp.com Nuclear 2oo Exelon Way Kennett Square, PA 19348 10CFR50, Appendix E February 5, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Peach Bottom Atomic Power Station, Units 2 & 3 Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 ERP-101, Revision 23, "Classification of Emergencies" ERP-1 01 Bases, Revision 3, "PBAPS EAL Technical Basis Manual"

Dear Sir/Madam:

Enclosed are revised Emergency Response Procedures (ERPs) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. These procedures are required to be submitted within thirty (30) days of their revision in accordance with 10CFR50, Appendix E, and 10CFR50.4.

Also, enclosed is a copy of a computer generated report index identifying the latest revisions of the PBAPS ERPs.

If you have any questions or require additional information, please do not hesitate to contact us.

Very truly yours, M. P. Gallagher Director - Licensing & Regulatory Affairs Mid-Atlantic Regional Operating Group Enclosures cc: H. J. Miller, Administrator, Region I, USNRC (2 copies)

A. C. McMurtray, USNRC Senior Resident Inspector, PBAPS

ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 & 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 EMERGENCY RESPONSE PROCEDURE ERP-101, "Classification of Emergencies" Revision 23 ERP-101 Bases, "PBAPS EAL Technical Basis Manual" Revision 3

Effective Date:- 02 ERP-101, Rev. 23 Page 1 of 32 RES/res PECO NUCLEAR PEACH BOTTOM UNITS 2 AND 3 EMERGENCY RESPONSE PROCEDURE ERP-101 CLASSIFICATION OF EMERGENCIES 1.0 RESPONSIBILITIES 1.1 Shift Management:

1.1.1 Recognize and classify an event or condition.

1.1.2 Assume duties of Emergency Director (ED).

1.2 Plant Manager or designated alternate:

1.2.1 Relieve acting ED.

1.2.2 Assume duties of ED.

2.0 INITIAL ACTIONS NOTE THE JUDGMENT OF THE EMERGENCY DIRECTOR TAKES PRECEDENCE OVER GUIDANCE IN THE PROCEDURE.

NOTE IDENTIFICATION AND CLASSIFICATION OF EMERGENCIES SHOULD BE ACCOMPLISHED WITHIN 15 MINUTES AFTER THE APPLICABLE EMERGENCY ACTION LEVELS (EALs) ARE MET.

2.1 Emergency Director shall:

2.1.1 Select categories appropriate for station events or conditions.

2.1.2 Review Emergency Action Level (EALs) for categories selected.

2.1.3 IF the event trigger is known to be spurious, THEN do not classify the event (i.e., false high reading, false radiation monitor readings, etc.)

2.1.4 Classify the event based on selected categories and most severe EALs.

2.1.5 IF the event or condition classifies as an emergency, THEN assume duties of ED and implement ERP-200.

ERP-101, Rev. 23 Page 2 of 32 RES/res 3.0 CONTINUING ACTIONS NOTE IT IS PREFERABLE TO OBTAIN EMERGENCY RESPONSE MANAGER (ERM)

CONCURRENCE PRIOR TO DE-ESCALATION.

3.1 IF emergency conditions dictate, THEN escalate or de escalate emergency classification.

4.0 FINAL CONDITIONS 4.1 Emergency conditions have been terminated, or ERP-C 1900, Recovery Phase Implementation has been implemented.

5.0 ATTACHMENTS AND APPENDICES 5.1 Attachment 1 - EAL Table of Contents and Tables 1 through 9. CM-I, CM-2, CM-3, CM-5 5.2 Attachment 2 - Terms and Definitions 6.0 SUPPORTING INFORMATION 6.1 Purpose 6.1.1 To provide the method for classifying an event or condition into one of four (4) emergency classifications described in the Nuclear Emergency Plan.

6.1.2 To provide pre-determined Protective Action Recommendations (PARs) for specific plant conditions whenever a General Emergency is declared.

6.2 Criteria For Use 6.2.1 Implement whenever conditions meet or exceed EALs listed in the Tables.

NOTE ISSUANCE OF A PAR REQUIRES A GENERAL EMERGENCY CLASSIFICATION AND CONVERSELY A GENERAL EMERGENCY CLASSIFICATION REQUIRES THE ISSUANCE OF A PAR.

6.2.2 PAR information in the tables, is expected to be used when an event rapidly progresses to a General Emergency or when the PAR is based only on plant conditions. Dose Assessment based PAR information may be obtained from the Dose Assessment Coordinator or the Dose Assessment Team Leader. In either case, the most conservative PAR available is to be used.

ERP-101, Rev. 23 Page 3 of 32 RES/res 6.2.3 Whenever the Emergency Operations Facility (EOF) is activated, then all PAR information from the ED should be submitted to the ERM.

CYL-4 6.3 Special Equipment None 6.4 References 6.4.1 EPA-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents 6.4.2 ERP-200, Emergency Director (ED) 6.4.3 ERP-C-1900, Recovery Phase Implementation 6.4.4 Nuclear Emergency Plan 6.4.5 NUMARC/NESP-007, Methodology for Development of Emergency Action Levels 6.4.6 NUREG 0654, FEMA-REP-1, Criteria for Preparations and Evaluation of Radiological Emergency Response Plans in Support of Nuclear Power Plants 6.4.7 PBAPS Technical Specifications 6.4.8 PBAPS Offsite Dose Calculation Manual 6.4.9 PBAPS Updated Final Safety Analysis Report 6.4.10 Reference Manual: Identification and Evaluation of Potentially Reportable Items 6.4.11 SE-l, Plant Shutdown from the Remote Shutdown Panel 6.4.12 SE-5, Earthquake 6.4.13 SE-10, Plant Shutdown from the Alternative Shutdown Panels 6.4.14 T-101, Reactor Pressure Vessel Control 6.4.15 T-102, Primary Containment Control 6.4.16 T-103, Secondary Containment Control 6.4.17 T-104, Radioactivity Release Control 6.4.18 T-116, RPV Flooding

ERP-101, Rev. 23 Page 4 of 32 RES/res 6.4.19 T-200, Primary Containment Venting 6.4.20 SO 67.7A, Verification of Suspected Earthquake or Seismic System Activation 6.4.21 US NRC Regulatory Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors 6.4.22 US NRC Response Technical Manual 6.5 Commitment Annotation 6.5.1 CM-i, NRC Inspection Report 50-277, 278/ 88 12/12 (T00349), (see Attachment 1, tables 1 through 9) 6.5.2 CM-2, Event INV Report 3-90-031, corrective action #7, (T00826), (see Attachment 1, table 1 for Reactor Fuel and table 3 for Fission Product Barrier) 6.5.3 CM-3, NRC URI 85-17-03, IN Inspection Report 86-06/06, (T01934), (see Attachment 1, table 9) 6.5.4 CM-4, Peach Bottom Inspection Report 92-19/19 (T02540), (see section 6.2.3) 6.5.5 CM-5, NRC Inspection 92-03/03, (T02541), (see Attachment 1, table 3 for Fission Product Barrier)

ERP-101, Rev. 23 Page 5 of 32 RES/res Attachment I EAL Table of Contents 1.0 Reactor Fuel 1.1 Coolant Activity ............................................................................................... 6 1.2 Irradiated Fuel or New Fuel ............................................................................. 7 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level ......................................................................................... 8 2.2 Reactor Power ................................................................................................. 9 3.0 Fission Product Barrier CM-2, CM-5 3.1 Initiating Condition Matrix .............................................................................. 10 3.2 Fission Product Barrier Table ......................................................................... 11 4.0 Secondary Containment Bypass 4.1 Main Steam Line ............................................................................................. 13 5.0 Radioactivity Release 5.1 Effluent Release and Dose ............................................................................ 14 5.2 In-Plant Radiation .......................................................................................... 16 6.0 Loss of Power 6.1 Loss of AC or DC Power ................................................................................ 18 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation .................................. 20 7.2 Loss of Decay Heat Removal Capabililty ...................................................... 21 7.3 Loss of Assessment/Communications Capabililty ......................................... 22 8.0 External Events 8.1 Security Events ............................................................................................. 24 8.2 Fire/Explosion and Toxic/Flammable Gases ...................................................... 26 8.3 Man-Made Events ......................................................................................... 28 8.4 Natural Events ............................................................................................... 29 9.0 Other CM-3 9 .1 G e n e ra l .............................................................................................................. 31 MODE 1 Run 2 Startup 3 Shutdown (hot) 4 Shutdown (cold) 5 Refueling D Defueled CM-1, All Tables

ERP-101, Rev. 23 Page 6 of 32 RES/res 1.0 Reactor Fuel 1.1 Coolant Activity CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Fuel Clad Degradation EVENT 1.1.1.a Applicable Modes: ALL Reactor Coolant activity > 4 pCi/gm Dose Equivalent Iodine 131 1.1.1.b Applicable Modes: 1, 2, 3 SJAE Discharge Radiation > 2.5x10 mR/hr ALERT None SITE AREA None EMERGENCY GENERAL None EMERGENCY

ERP-101, Rev. 23 Page 7 of 32 RES/res 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Unexpected Rise in Plant Radiation or Airborne Concentration.

EVENT 1.2.1.a Applicable Modes: ALL Uncontrolled water level drop in the spent fuel pool with all irradiated fuel assemblies remaining covered by water 1.2.1.b Applicable Modes: ALL Unexpected Skimmer Surge Tank low level alarm AND Visual observation of an uncontrolled water level drop below the fuel pool skimmer surge tank inlet IC Unexpected Rise in Plant Radiation 1.2.1.c Applicable Modes: ALL Radiological readings exceed 600 mR/hr one foot away OR 1200 mR/hr at the external surface of any dry storage system ALERT IC Major Damage to Irradiated Fuel, or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel 1.2.2.a Applicable Modes: ALL Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1) 1.2.2.b Applicable Modes: ALL Report of visual observation of irradiated fuel uncovered 1.2.2.c Applicable Modes: 5 (With Reactor Refueling Cavity Flooded)

Water Level < 458" above RPV instrument zero for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering 1.2.2.d Applicable Modes: ALL Water Level < 232ft 3 inches plant elevation for the Spent Fuel Pool that will result in Irradiated Fuel uncovering SITE AREA None EMERGENCY GENERAL None EMERGENCY I Table 1-1 Refuel Floor ARMs 3-7 (7-9) Steam Separator Pool 3-8 (7-10) Refuel Slot 3-9(7-11) Fuel Pool 3-10(7-12) Refueling Bridge

ERP-101, Rev. 23 Page 8 of 32 RES/res 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Reactor Coolant System Leakage EVENT 2.1.1 Applicable Modes: 1, 2, 3, 4 The following conditions exist:

Unidentified Primary System Leakage > 10 gpm into the Drywell OR Identified Primary System Leakage > 25 gpm into the Drywell ALERT None SITE AREA IC Loss of Water Level in the Reactor Vessel That Has or Will Uncover fuel in EMERGENCY the Reactor Vessel 2.1.3 Applicable Modes: 4, 5 RPV level < -172" GENERAL None EMERGENCY

ERP-101, Rev. 23 Page 9 of 32 RES/res 2.0 Reactor Pressure Vessel 2.2 Reactor Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL None EVENT ALERT IC Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful 2.2.2 Applicable Modes: 1, 2 Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS SCRAM to make Reactor shutdown SITE AREA IC Failure of Reactor Protection System Instrumentation to Complete or Initiate EMERGENCY an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful 2.2.3 Applicable Modes: 1, 2 RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

GENERAL IC Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core 2.2.4 Applicable Modes: 1, 2 RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

AND Torus Temperature is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-1 02, T/T-1) OR RPV level <-200"

• • • PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

ERP-101, Rev. 23 Page 10 of 32 RES/res 3.0 Fission Product Barrier Table 3.1 Initiating Condition Matrix USE TABLE 3.2, "FISSION PRODUCT BARRIER STATUS TABLE" FOR CLASSIFYING EVENT CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL 3.1.1 Applicable Modes: 1, 2, 3 EVENT ANY Loss OR ANY Potential Loss of Primary Containment ALERT 3.1.2 Applicable Modes: 1, 2, 3 ANY Loss OR ANY Potential Loss of EITHER Fuel Clad OR RCS SITE AREA 3.1.3 Applicable Modes: 1, 2, 3 EMERGENCY Loss of OBOTH Fuel Clad AND RCS OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS, AND Loss of ANY Additional Barrier GENERAL 3.1.4 Applicable Modes: 1, 2, 3 EMERGENCYAN Loss of ANY Two Barhiers AND Potential Loss of Third Barrier

• • • PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles. (See Fission Product Barrier Table 3.2 for exception based on extremely Hi Containment Radiation Levels.)

NOTES:

1. If a "Loss" condition is satisfied, the "Potential Loss" category can be considered satisfied. This is accounted for in the matrix contained in the Fission Product Barrier Table 3.2 used to determine the proper classification based on Fission Product Barrier status.

2. For all conditions listed in Fission Product Barrier Table 3.2, the barrier failure column is only satisfied if it fails when called upon to mitigate an accident. For example, failure of both containment isolation valves to isolate with a downstream pathway to the environment is only a concern during an accident.

If this condition exists during normal power operations, it will be an active Technical Specification Action Statement. However, during accident conditions, this will represent a breach of containment.

/'

ERP-101, Rev. 23 3.2 Fission Product Barrier Status Table Page 11 of 32 Applicable Modes: 1, 2, 3 RES/res

- T -

Barrier I-uel Clad Reactor Coolant System Primary Containment Parameter Reactor Coolant Activity RPV Level RPV Level Unknown RCS Leak Rate Drywell Pressure Drywell Radiation

ERP-101, Rev. 23 3.2 Fission Product Barrier Status Table Page 12 of 32 Applicable Modes: 1, 2. 3 I -Ir Reactor Coolant System Primary Containment s Potential Loss Loss I Loss Unisolable primary system leakage outside drywell as Failure of both valves inany indicated by T-1 03, one line to close AND Temperature Action Level downstream pathway to the is exceeded in ONE area environment exists requiring a SCRAM OR OR Unisolable primary system Intentional venting per leakage outside drywell as T-200 is required indicated by T-1 03, Radiation Action Level is OR exceeded in ONE area Unisolable primary system leakage outside drywell as requiring a SCRAM indicated by T-1 03, Temperature Action Level is exceeded in ONE area requiring a SCRAM OR Unisolable primary system leakage outside drywell as indicated by a T-103, Radiation Action Level is exceeded in ONE area requiring a SCRAM Ilry ,Vl IILIUll III I1l JUUy9IJllitl Ul LIV Fi*ulylrluy I Ulfe[uLU Any condition in the judgment of the Emergency Director Any condition in the judgment of the Emergency Dir that indicates Loss or Potential Loss of the FUEL CLAD that indicates Loss or Potential Loss of the RCS barrier that indicates Loss or Potential Loss of the Primary barrier Containment barrier In the table below, circle all of the appropriate X's in each applicable row for each Loss or Potential Loss of Fission Product Barrier as determined by the table above.

Classify the event as identified in the table heading if all X's in a column under that heading are circled.

Fission Product Barrier Status Unusual ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Event Fuel Clad - Loss X X X XX X X X Fuel Clad - Potential Loss X X X X X Reactor Coolant System - Loss _ X X X X X X Reactor Coolant System-Potential Loss X X X X X Primary Containment - Loss X X X IX X Primary Containment - Potential Loss X X Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles. (Upgrade PAR for D/W Rad > 6x10W'Rhi)

ERP-101, Rev. 23 Page 13 of 32 RES/res 4.0 Secondary Containment Bypass 4.1 Main Steam Line CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Fuel Clad Degradation EVENT 4.1.1 Applicable Modes: 1, 2, 3 Main Steam Line HiHi Radiation (1OxNFPB)

ALERT IC RCS Leak Rate 4.1.2 Applicable Modes: 1, 2, 3 Indication of a Main Steam Line Break:

Hi Steam Flow Annunciator AND Hi Steam Tunnel Temperature Annunciator OR Direct report of steam release SSITE AREA None EMERGENCY GENERAL None

,EMERGENCY

ERP-101, Rev. 23 Page 14 of 32 RES/res 5.0 Radioactivity Release 5.1 Effluent Release and Dose CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the EVENT Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer 5.1.1.a Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds TWO TIMES the HiHi alarm setpoint value for > 60 minutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate using computer dose model exceeds 0.114 mRem/hr TPARD OR 0.342 mRem/hr child thyroid CDE based on a 60 minute average Note: If the required dose projections cannot be completed within the 60 minute period, then the declaration must be made based on the valid sustained monitor reading.

5.1.1.b Applicable Modes: ALL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO TIMES Tech Specs (Liquid Release ODCM 3.8.B.1 and Gaseous Release ODCM 3.8.C.1.b) for

> 60 minutes ALERT IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer 5.1.2.a Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds TWO HUNDRED TIMES the HiHi alarm setpoint value for > 15 minutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mRemlhr TPARD OR 34.2 mRemlhr child thyroid CDE based on a 15 minute average Note: Ifthe required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

5.1.2.b Applicable Modes: ALL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO HUNDRED TIMES Tech Specs (Liquid Release ODCM 3.8.B.1 and Gaseous Release ODCM 3.8.C.l.b) for

> 15 minutes

ERP-101, Rev. 23 Page 15 of 32 RES/res SITE AREA IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the EMERGENCY Actual or Projected Duration of the Release 5.1.3 Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:

Main Stack 5.84 jLCi/cc Vent Stack 2.08E-3 gtCi/cc Torus Vent 203 cpm Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

OR Projected offsite dose using computer dose model exceeds 100 mRem TPARD OR 500 mRem child thyroid CDE OR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mRem/hr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 500 mRem for one hour of inhalation 1

GENERAL IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid EMERGENCY for the Actual or Projected Duration of the Release Using Actual Meteorology 5.1.4 Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:

Main Stack 58.4 [lCi/cc Vent Stack 2.08E-2 pCi/cc Torus Vent 2000 cpm Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

OR Projected offsite dose using computer dose model exceeds 1000 mRem TPARD OR 5000 mRem child thyroid CDE OR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mRemlhr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mRem for one hour of inhalation

• • • PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

NOTE: CDE = Committed Dose Equivalent, TPARD = Total Protective Action Recommendation Dose

ERP-101, Rev. 23 Page 16 of 32 RES/res 5.0 Radioactivity Release 5.2 In-Plant Radiation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Unexpected Rise in Plant Radiation or Airborne Concentration EVENT 5.2.1 Applicable Modes: ALL Valid Direct Area Radiation Monitor readings rise by a factor of 1000 over normal*

levels Normal levels can be considered as the highest reading in the past twenty four hours excluding the current peak value.

ALERT IC Release of Radioactive Material or Rises in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown 5.2.2.a Applicable Modes: ALL Valid radiation level readings > 5000 mR/hr in areas requiring infrequent access to maintain plant safety functions as identified in procedure SE-1, SE-10 AND Access is required for safe plant operation, but is impeded, due to radiation dose rates 5.2.2.b Applicable Modes: ALL Valid Control Room OR Central Alarm Station radiation reading > 15 mR/hr SITE AREA None EMERGENCY GENERAL None EMERGENCY

ERP-101, Rev. 23 Page 17 of 32 RES/res This Intentionally Left Blank

ERP-101, Rev. 23 Page 18 of 32 RES/res 6.0 Loss of Power 6.1 Loss of AC or DC Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT 6.1.1.a Applicable Modes: ALL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND At least Two Diesel Generators are supplying power to their respective 4 KV emergency busses IC Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes 6.1.1.b Applicable Modes: 4, 5 Unplanned Loss of ALL safety related DC Power indicated by

< 107.5 VDC on DC Panels 2(3)0D21, 22, 23, 24 for >15 minutes ALERT IC AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout 6.1.2.a Applicable Modes: 1, 2, 3 The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND Only One 4 KV emergency bus powered from a Single Onsite Power Source due to the Loss of: Three of Four Division Diesel Generators, D/G Output Breakers, or 4 KV Emergency Busses as indicated by bus voltage IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode 6.1.2.b Applicable Modes: 4, 5, D The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power

ERP-101, Rev. 23 Page 19 of 32 RES/res SITE AREA IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EMERGENCY 6.1.3.a Applicable Modes: 1, 2, 3 The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC IC Loss of All Vital DC Power 6.1.3.b Applicable Modes: 1, 2, 3 Loss of ALL Safety Related DC Power indicated by < 107.5 VDC on DC Panels 2(3)0D21, 22, 23, 24 for > 15 minutes GENERAL IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EMERGENCY 6.1.4 Applicable Modes: 1, 2, 3 Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure of ALL Emergency Diesel Generators to supply power to 4 KV emergency busses AND At least one of the following conditions exist:

"* Restoration of at least One emergency bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is NOT likely OR

"* Reactor Water Level cannot be maintained > -172 OR

"* Torus temperature is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-102, T/IT-1)

• • • PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

ERP-101, Rev. 23 Page 20 of 32 RES/res 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Inability to Reach Required Shutdown Mode Within Technical Specification EVENT Limits 7.1.1 Applicable Modes: 1, 2, 3 Inability to reach required shutdown mode within Tech. Spec. LCO required action completion time.

ALERT IC Control Room Evacuation Has Been Initiated 7.1.2 Applicable Modes: ALL Entry into SE-1 or SE-10 procedure for Control Room evacuation SITE AREA IC Control Room Evacuation Has Been Initiated and Plant Control Cannot Be EMERGENCY Established 7.1.3 Applicable Modes: ALL The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-I or SE-1 0 within 15 minutes GENERAL None EMERGENCY

ERP-101, Rev. 23 Page 21 of 32 RES/res 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL None EVENT ALERT IC Inability to Maintain Plant in Cold Shutdown 7.2.2 Applicable Modes: 4, 5 The following conditions exist:

Unplanned Loss of ALL Tech Spec required systems available to provide Decay Heat Removal functions AND Uncontrolled Temperature rise that either:

• Exceeds 212 'F (Excluding a <15 minute rise >2120 F with a heat removal function restored)

OR

• Results in temperature rise approaching 212 'F (with NO heat removal function restored)

SITE AREA IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EMERGENCY 7.2.3 Applicable Modes: 1, 2, 3 Loss of TORUS heat sink capabilities as evidenced by T-102 T/T legs directing a T 112 Emergency Blowdown GENERAL None EMERGENCY

ERP-1 01, Rev. 23 Page 22 of 32 RES/res 7.0 Internal Events 7.3 Loss of Assessment I Communication Capability CLASSIFICATION EMERGENCY ACTION LEVEL 4

IC Unplanned Loss of Most or All Safety System Annunciation or Indication in UNUSUAL The Control Room for Greater Than 15 Minutes EVENT 7.3.1 .a Applicable Modes: 1, 2, 3 Unplanned loss of most or all safety system annunciators (Table 7-1) OR indicators (Table 7-2) for> 15 minutes requiring increased surveillance to safely operate the unit(s).

IC Unplanned Loss of All Onsite or Offsite Communications Capabilities 7.3.1 .b Applicable Modes: ALL Loss of ALL Onsite communications (Table 7-3) affecting the ability to perform routine operations OR Loss of ALL Offsite communications (Table 7-3)

ALERT IC Unplanned Loss of Most or All Safety System Annunciation or Indication In Control Room With Either (1) a Significant Transient in Progress, or (2)

Compensatory Non-Alarming Indicators are Unavailable 7.3.2 Applicable Modes: 1, 2, 3 Unplanned loss of most or all safety system annunciators (Table 7-1) OR indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit(s)

AND EITHER A significant plant transient is in progress (Table 7-4) OR the plant monitoring system (PMS) is unavailable.

SITE AREA IC Inability to Monitor a Significant Transient in Progress EMERGENCY 7.3.3 Applicable Modes: 1, 2, 3 Loss of safety system annunciators (Table 7-1)

AND indicators (Table 7-2)

AND PMS AND a significant plant transient is in progress. (Table 7-4)

GENERAL None

- EMERGENCY

ERP-101, Rev. 23 Page 23 of 32 RES/res Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions Table 7-3 Communications Onsite Offsite Site Phones (GTE System) X X OMNI System X X Plant Public Address X Station Radio X NRC (FTS-2000) X PA State Police Radio X Load Dispatcher Radio X PECO Dial Network X Table 7-4 Significant Plant Transients SCRAM Recirc Runbacks > 25% thermal power Sustained power oscillations 25% peak to peak Stuck open relief valve(s)

ECCS injection

ERP-101, Rev. 23 Page 24 of 32 RES/res 8.0 External Events 8.1 Security Threats I

CLASSIFICATION EMERGENCY ACTION LEVEL IC Confirmed Security Event Which Indicates a Potential Degradation in the UNUSUAL Level of Safety of the Plant EVENT 8.1.1 Applicable Modes: ALL A credible threat to the station reported by the NRC.

OR An actual threat that meets ALL of the following criteria:

"* A credible threat reported by any other outside agency or determined per SY-AA-101-132; AND

"* Is specifically directed towards the station; AND

"* Is imminent (within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).

OR Attempted intrusion and attack to the Protected Areas OR Attempted sabotage discovered within the Protected Areas OR Hostage/Extortion situation that threatens normal plant operations ALERT IC Security Event in a Plant Protected Area 8.1.2 Applicable Modes: ALL Intrusion into plant protected areas by a hostile force OR Confirmed bomb, sabotage or sabotage device discovered in the Protected Areas SITE AREA IC Security Event in a Plant Vital Area EMERGENCY 8.1.3 Applicable Modes: ALL Intrusion into plant Vital area by a hostile force OR Confirmed bomb, sabotage or sabotage device discovered in a Vital Area GENERAL IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold EMERGENCY Shutdown 8.1.4 Applicable Modes: ALL Loss of physical control of the control room due to security event OR Loss of physical control of all remote shutdown capability due to security event "PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

ERP-101, Rev. 23 Page 25 of 32 RES/res This Page Intentionally Left Blank

ERP-101, Rev. 23 Page 26 of 32 RES/res 8.0 External Events 8.2 Fire / Explosion and Toxic I Flammable Gases CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of EVENT Detection 8.2.1.a Applicable Modes: ALL Fire within ON-1 14 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant 8.2.1.b Applicable Modes: ALL Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant OR Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event IC Natural and Destructive Phenomena Affecting the Protected Area 8.2.1.c Applicable Modes: ALL Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment ALERT IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown 8.2.2.a Applicable Modes: ALL The following conditions exist:

Fire or explosion which potentially makes inoperable:

Two or More subsystems of a Safe Shutdown System (Table 8-2) OR Two or More Safe Shutdown Systems OR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Mode

ERP-101, Rev. 23 Page 27 of 32 RES/res IC Release of Toxic or Flammable Gases Within a Facility Structure Which ALERT Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown 8.2.2.b Applicable Modes: ALL Report or detection of toxic gases within Plant Vital Structures (Table 8-1) in concentrations that will be life threatening to plant personnel OR Report or detection of flammable gases within Plant Vital Structures (Table 8-1) in concentrations affecting the safe operation of the plant SITE AREA None EMERGENCY GENERAL None EMERGENCY Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower Table 8-2 Safe Shutdown Systems Diesel Generators 4KV Safeguard Buses ADS HPCI RCIC RHR (All Modes)

Core Spray HPSW ESW SBGTS ECW CAC/CAD PCIS Control Room Ventilation

ERP-101, Rev. 23 Page 28 of 32 R ES/res 8.0 External Events 8.3 Man-Made Events CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Destructive Phenomena Affecting the Protected Area EVENT 8.3.1.a Applicable Modes: ALL Vehicle crash within protected area boundary that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.

8.3.1.b Applicable Modes: ALL Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

ALERT IC Destructive Phenomena Affecting the Plant Vital Area 8.3.2 Applicable Modes: ALL Vehicle crash affecting Plant Vital Structures (Table 8-1)

OR Turbine failure generated missiles result in any visible structural damage to or penetration of any Plant Vital Structures (Table 8-1)

SITE AREA None EMERGENCY GENERAL None EMERGENCY Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower

ERP-101, Rev. 23 Page 29 of 32 RES/res 8.0 External Events 8.4 Natural Events CLASSIFICATION EMERGENCY ACTION LEVEL IC Natural and Destructive Phenomena Affecting the Protected Area UNUSUAL EVENT 8.4.1.a Applicable Modes: ALL Earthquake >.01 g as determined by procedure SO 67.7.A 8.4.1.b Applicable Modes: ALL Report by plant personnel of tornado striking within protected areas OR Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes 8.4.1.c Applicable Modes: ALL Assessment by the control room that an event has occurred. (Natural and Destructive Phenomena Affecting the Protected Areas) 8.4.1.d Applicable Modes: All High River level > 112' OR Low River level < 98.5' IC Natural and Destructive Phenomena Affecting the Plant Vital Area ALERT 8.4.2.a Applicable Modes: ALL Earthquake >.05 g (Operating Basis Earthquake OBE) as determined by procedure SO 67.7.A 8.4.2.b Applicable Modes: ALL Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 8-1) 8.4.2.c Applicable Modes: ALL Report of any visible structural damage to any Plant Vital Structure (Table 8-1) 8.4.2.d Applicable Modes: All High River level > 116' OR Low River level < 92.5'

-I-SITE AREA None EMERGENCY 4-GENERAL None EMERGENCY

ERP-101, Rev. 23 Page 30 of 32 RES/res fable 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower

ERP-101, Rev. 23 Page 31 of 32 RES/res 9.0 Other 9.1 General CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Other Conditions Existing Which in the Judgment of the Emergency Director EVENT Warrant Declaration of an Unusual Event 9.1.1 Applicable Modes: ALL Other conditions exist which in the judgment of the Emergency Director indicate a potential degradation of the level of safety of the plant ALERT IC Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert 9.1.2 Applicable Modes: ALL Other conditions exist which in the Judgment of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted SITE AREA IC Other Conditions Existing Which in the Judgment of the Emergency Director EMERGENCY Warrant Declaration of Site Area Emergency 9.1.3 Applicable Modes: ALL Other conditions exist which in the Judgment of the Emergency Director indicate actual or likely major failures of plant functions needed for protection of the public GENERAL IC Other Conditions Existing Which in the Judgment of the Emergency Director EMERGENCY Warrant Declaration of General Emergency 9.1.4 Applicable Modes: ALL Other conditions exist which in the Judgment of the Emergency Director indicate:

(1) actual or imminent substantial core degradation with potential for loss of containment, or (2) potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary

• • • PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

S!1,

( Rev. 23

,ge 32 of 32 RES/res Attachment 2 TERMS AND DEFINII Plant parameters or other condition Events in progress or have occurred, that indicate a which if met or exceeded the emergency potential degradation of the level of safety of the plant.

classification level and requires a No releases of radioactive material requiring off-site declaration of emergency. response or monitoring are expected unless further degradation of safety systems occurs.

System, subsystem, train, component, or device, and all auxiliaries required for their operation, is capable of performing its specified function in the intended manner.

Recommendation made to the state action Events in progress or have occurred that involve actual or to be taken to avoid or reduce potential substantial degradation of the level of safety of projected dose to the public. the plant. Any releases of radioactive material are expected to be limited to small fractions of the Environmental Protective Agency (EPA) Protective Action Guidelines (PAG) exposure levels.

An estimate of radiation dose which affected individuals could potentially receive if protective actions are not taken.

Total Protective Action Recommendation Dose. (TPARD = External Dose & Inter nal Dose & Dose Due to 4-Day Shine)

Committed Dose Equivalent. (CDE = in ternal Organ Dose from Ingestion)

Committed Effective Dose Equivalent.

(CEDE = Internal Whole Body Dose from Ingestion)

Total Effective Dose Equivalent. (TEDE

= Deep Dose Equivalent & CEDE Dose)

Action guidelines based on projections Events in progress or which have occurred that involve actual for the total integrated dose a member or likely major failures of plant functions needed for of the public would receive for the protection of the public. Any releases of radioactive duration of the emergency. material are not expected to exceed EPA PAG exposure levels except near site boundary.

An act conducted by a person or persons with the intent of damaging or impairing the operation of the plant.

A security threat as illustrated by Events in progress or which have occurred that involve actual attempted entry or sabotage with the or imminent substantial core degradation or melting with intent to gain physical control of the potential for loss of containment integrity. Releases of plant. radioactive material can be reasonably expected to exceed EPA PAG exposure levels off-site for more than the immediate site area.

2/8 /.02 ERP-101 BASES, Rev 3 I?1 Page 1of 133 J DA/Idt PBAPS EAL Technical Basis Manual Table of ContentsSection I - Introduction ...................................................................................................... 2 Section II - Acronyms ............................................................................................................... 5 Section III - EAL Technical Basis ........................................................................................ 8 1.0 Reactor Fuel 1.1 Coolant Activity ............................................................................................... 9 1.2 Irradiated Fuel or New Fuel ........................................................................... 11 2.0 Reactor Pressure Vessel 2.1 Reactor W ater Level ..................................................................................... 23 2.2 Reactor Power ............................................................................................... 27 3.0 Fission Product Barrier 3.1 Initiating Condition Matrix ............................................................................. 33 3.2 Fuel Clad Barrier Thresholds ......................................................................... 36 3.3 Reactor Coolant System Barrier Thresholds ................................................ 43 3.4 Primary Containment Barrier Thresholds ...................................................... 50 3.5 Fission Product Barrier Table ....................................................................... 61 4.0 Secondary Containment Bypass 4.1 Main Steam Line ............................................................................................ 63 5.0 Radioactivity Release 5.1 Effluent Release and Dose .......................................................................... 65 5.2 In-Plant Radiation .......................................................................................... 75 6.0 Loss of Power 6.1 Loss of AC or DC Power ............................................................................... 81 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation ................................... 91 7.2 Loss of Decay Heat Removal Capability ....................................................... 94 7.3 Loss of Assessment/Communications Capability .......................................... 98 8.0 External Events 8.1 Security Events ............................................................................................... 105 8.2 Fire/Explosion and Toxic/Flammable Gases ............................. 110 8.3 Man-Made Events ........................................................................................... 117 8.4 Natural Events ................................................................................................. 120 9.0 Other 9 .1 G e n e ra l ........................................................................................................... 1 29

ERP-101 BASES, Rev 3 Page 2 of 133 Section I - Introduction This manual contains the technical basis for the Emergency Action Levels as utilized in ERP 101, Classification of Emergencies. The format and use of this manual is as follows.

1. Heading and Sub-Heading There are nine major headings each containing one or more sub-headings. These are as follows:

1.0 Reactor Fuel 1.1 Coolant Activity 1.2 Irradiated Fuel or New Fuel 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level 2.2 Reactor Power 3.0 Fission Product Barrier 3.1 Initiating Condition Matrix 3.2 Fuel Clad Barrier Thresholds 3.3 Reactor Coolant System Barrier Thresholds 3.4 Primary Containment Barrier Thresholds 3.5 Fission Product Barrier Table 4.0 Secondary Containment Bypass 4.1 Main Steam Line 5.0 Radioactivity Release 5.1 Effluent Release and Dose 5.2 In-Plant Radiation 6.0 Loss of Power 6.1 Loss of AC or DC Power 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation 7.2 Loss of Decay Heat Removal Capability 7.3 Loss of Assessment/Communications Capability 8.0 External Events 8.1 Security Events 8.2 Fire/Explosion and Toxic/Flammable Gases 8.3 Man-Made Events 8.4 Natural Events 9.0 Other 9.1 General

ERP-101 BASES, Rev 3 Page 3 of 133

2. Emergency Classification Level and Number Identification The classifications range from Unusual Event through Alert, Site Area Emergency to General Emergency. For each sub-heading, there may not be an EAL in every classification level. Each EAL is individually and uniquely numbered. No two numbers are the same.

3. INITIATING CONDITION The Initiating Condition or IC (as described in NUMARC NESP-007) is contained in this section. ICs are a predetermined subset of conditions where either the potential exists for a radiological emergency or such an emergency has occurred. Additionally, ICs are the means by which EALs for different nuclear power plants are standardized.

4. EAL Each Emergency Action Level exactly as it is contained in ERP-101.

5. MODE The mode that the EAL is applicable in is contained here. There are six MODEs (1, 2, 3, 4 and 5 and defueled) that are used. PBAPS also uses mode switch position.

These positions are stated below and are Run, Startup, Shutdown and Refueling. It should be noted that these MODEs are entry level conditions. The EAL is applicable if the plant was in the MODE at the start of the event. Subsequent positions of the mode selector switch should be ignored for purposes of classification.

MODE MODE SWITCH POSITION 1 Run 2 Startup 3 Shutdown (hot) 4 Shutdown (cold) 5 Refueling D N/A (defueled)

6. BASIS The technical basis of each EAL is contained in this section. This includes any necessary calculations and also includes escalation references.

7. DEVIATION Any deviations from the NUMARC NESP-007 methodology are contained in this section. If there are no deviations, NONE is used.

8. REFERENCES All applicable references used in developing the technical basis for each EAL are contained in this section.

9. GENERAL EAL IMPLEMENTATION PHILOSOPHY The following guidance is provided to describe the philosophy used in the implementation of ERP-1 01 by the Emergency Director (ED) in making emergency classifications. CM-1 (ERP-101)

ERP-101 BASES, Rev 3 Page 4 of 133 In most cases, the emergency classification process is a straight-forward comparison of important plant parameters to the emergency action levels (EAL's). The instruments and annunciators referred to in the Emergency Classification Tables are presented as primary indicators and should be validated by plant conditions or event conditions.

A broad spectrum of discretion in classifying events is provided to the ED under the "General Conditions" category. In using the "General Conditions" category and in classifying emergencies under circumstances which are not straight-forward use of the EAL's, the ED should be mindful than an approach is needed which is conservative with respect to public, plant, and personnel safety and with respect to ensuring the adequacy of personnel and technical support. Conservative decisions must be made if the ED has any doubt regarding the health and safety of the public.

The ED should be mindful that declaring Unusual Events provide the Company and off-site agencies the opportunity for early information regarding the event and for early activation of resources and may be considered a "no consequence decision."

Conversely, not declaring an Unusual Event when there is credible (but, not clear) bases for doing so, would appear to be less than open or candid and could have serious adverse consequences. Although the consequences of declaring an Unusual Event are limited, inappropriate classifications do not accurately indicate the significance of the event to the public and emergency responders and should be avoided.

At the Alert, Site Area and General Emergency levels, clearly the threat to the plant and to the public is at a heightened level. Rapid application of resources and preparation for providing for the public health and safety are appropriate. Because of the magnitude of resource mobilization and the potential disruption of normal public activities, an overly conservative or an inappropriately early declaration of these levels is not advisable.

Events that meet the Emergency Action Level criteria for event declaration, but which are terminated before they are identified and declared, should still be classified and reported, but not declared to implement the Emergency Plan.

All EAL's may not consider trends, rates of change, or status changes in equipment availability. In the event of rapidly changing parameters trending toward an increased emergency classification, the ED can appropriately decide that the higher level EAL will be exceeded and escalate the classification early. In the event of trends toward a decreased emergency classification, parameter values must be below the EAL's to de-escalate.

In the event of a "spike" which rapidly exceeds and then decreases below an EAL, entry into the Emergency Plan or escalation to the higher classification "in retrospect" is not appropriate unless the "spike" is indicative of continuing degrading conditions which will lead to an escalated emergency classification level. This statement does not apply if the EAL includes a "spike". Spurious alarms or parameters which are known to be invalid indicators of actual plant conditions or of the emergency classification, should not be used to declare emergency classifications.

ERP-101 BASES, Rev 3 Page 5 of 133 Section II- Acronyms AC - Alternating Current ADS - Automatic Depressurization System APRM - Average Power Range Monitor ARI - Alternate Rod Insertion ARM - Area Radiation Monitor ATWS - Anticipated Transient Without Scram BRP - Bureau of Radiation Protection CAC - Containment Atmosphere Control CAD - Containment Atmosphere Dilution CDE - Committed Dose Equivalent CFM - Cubic Feet Per Minute CFR - Code of Federal Regulations CRD - Control Rod Drive CS - Core Spray DBA - Design Basis Accident DC - Direct Current DEI - Dose Equivalent Iodine EAL - Emergency Action Level ECCS - Emergency Core Cooling Systems ECW - Emergency Cooling Water EDG - Emergency Diesel Generator EPA - Environmental Protection Agency ERP-C - Emergency Response Procedure - Common ESW - Emergency Service Water FC - Fuel Clad (Barrier)

FTS - Federal Telephone System GPM - Gallons Per Minute HCTL - Heat Capacity Temperature Limit HPCI - High Pressure Coolant Injection HPSW - High Pressure Service Water IC - Initiating Condition IRM - Intermediate Range Monitor KV - KiloVolt LCO - Limiting Condition for Operation LOCA - Loss of Coolant Accident LPCI - Low Pressure Coolant Injection MPH - Miles Per Hour mR/hr - Milli Roentgen Per Hour MSIV - Main Steam Isolation Valve NFPB - Normal Full Power Background NPSH - Net Positive Suction Head NRC - Nuclear Regulatory Commission NUMARC - Nuclear Management and Resources Council ODCM - Offsite Dose Calculation Manual OPCON - Operating Condition PBAPS - Peach Bottom Atomic Power Station PEMA - Pennsylvania Emergency Management Agency PC - Primary Containment (Barrier)

PCIS - Primary Containment Isolation System PSIG Pounds Square Inch Gauge RC Reactor Coolant (Barrier)

ERP-101 BASES, Rev 3 Page 6 of 133 RCIC - Reactor Core Isolation Cooling RCS - Reactor Coolant System RHR - Residual Heat Removal RPS - Reactor Protection System RPV - Reactor Pressure Vessel SBGTS - Standby Gas Treatment System SBO - Station Blackout SJAE - Steam Jet Air Ejector SRM - Source Range Monitor SRV - Safety Relief Valve TAF - Top of Active Fuel TPARD - Total Protective Action Recommendation Dose TRIPs - Transient Response Implementation Plan Procedures p,Ci/cc - Micro Curie Per Cubic Centimeter pCi/gm - Micro Curie Per Gram UFSAR - Updated Final Safety Analysis Report VDC - Volts Direct Current

ERP-101 BASES, Rev 3 Page 7 of 133 This page intentionally left blank.

ERP-101 BASES, Rev 3 Page 8 of 133 Section III - EAL Technical Basis

ERP-101 BASES, Rev 3 Page 9 of 133 1.0 Reactor Fuel 1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.a IC Fuel Clad Degradation EAL Reactor Coolant activity > 4 pCilgm Dose Equivalent Iodine 131 MODE All BASIS Coolant activity in excess of Technical Specifications (> 4 4iCi/gm) is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition. This level is chosen to be above any possible short duration spikes under normal conditions. An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample (as determined by laboratory confirmation).

However, fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another cause is known, e.g., Reactor Coolant System chemical decontamination evolution (during shutdown) is ongoing with resulting high activity levels.

This event will be escalated to an Alert when Reactor Coolant activity exceeds 300 4Ci/gm Dose Equivalent Iodine 131 per Fission Product Barrier Table.

DEVIATION None REFERENCES Technical Specification Section 3.6.B NUMARC NESP-007, SU4.2

ERP-101 BASES, Rev 3 Page 10 of 133 1.0 Reactor Fuel 1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.b IC Fuel Clad Degradation EAL SJAE Discharge Radiation > 2.5x103 mR/hr MODE 1, 2,3 BASIS The steam jet air ejector discharge (Offgas) radiation monitor RR-2(3)-17-152 in the Control Room would be one of the first indicators of a degrading core. The high-high alarm is set at the Technical Specification limit of 2.5x10 3 mR/hr. This instrument takes a sample before the recombiner. This indicator of elevated activity is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition.

Escalation of this IC to the Alert level is via the Fission Product Barrier Degradation Monitoring ICs.

DEVIATION The MODE applicability [1,2,31 is a deviation from NUMARC [all] in that the SJAE Radiation Monitor and Main Steam Line Radiation Monitors will only be a valid indication of Fuel Clad Degradation in those MODE's. At Peach Bottom, there are no other monitors which can be an indicator of Fuel Clad Degradation. Degradation in cold shutdown or refueling will be first indicated by ventilation release monitor's which are covered by EAL on Effluent Release and Dose.

REFERENCES Technical Specifications Section 3.8.C.7.a NUMARC NESP-007, SU4.1

ERP-101 BASES, Rev 3 Page 11 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.2.1.a IC Unexpected Rise in Plant Radiation or Airborne Concentration.

EAL Uncontrolled water level drop in the spent fuel pool with all irradiated fuel assemblies remaining covered by water MODE All BASIS UNCONTROLLED - An unexplained level drop that cannot be quickly terminated and is not the result of a planned evolution.

This event tends to have a long lead time relative to potential for radiological release outside the site boundary, thus impact to public health and safety is very low.

In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these types of events via this EAL is appropriate given their potential for increased doses to plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious event.

This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.

DEVIATION None REFERENCES NUMARC NESP-007, AU2.2 Technical Specifications

ERP-101 BASES, Rev 3 Page 12 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.2.1.b IC Unexpected Rise in Plant Radiation or Airborne Concentration.

EAL Unexpected Skimmer Surge Tank low level alarm AND Visual observation of an uncontrolled water level drop below the fuel pool skimmer surge tank inlet MODE All BASIS UNEXPECTED - An alarm that is not a result of a planned evolution.

UNCONTROLLED - An unexplained level drop that cannot be quickly terminated and is not the result of a planned evolution.

A drop in the Spent Fuel Pool level or the RPV [when in refueling and flooded up with the gates removed] will result in a control room annunciator Fuel Pool Cooling and Cleanup System Trouble Alarm. This Control Room alarm directs an operator to be dispatched to a local alarm panel which will identify the Skimmer Surge Tank low level alarm. This alarm is validated with visual observation of a decreasing Spent Fuel Pool level. If the spent fuel pool level decreases below the inlet to the skimmer surge tank, without a planned event such as removing a large piece of equipment, there must be a leak in the spent fuel pool or the RPV.

This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event.

In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these types of events via this EAL is appropriate given their potential for increased doses to plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious event.

This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.

ERP-101 BASES, Rev 3 Page 13 of 133 DEVIATION None REFERENCES NUMARC NESP-007, AU2.1

ERP-101 BASES, Rev 3 Page 14 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.2.1.c IC Unexpected Rise in Plant Radiation EAL Radiological readings exceed 600 mRPhr one foot away OR 1200 mR/hr at the external surface of any dry storage system MODE All BASIS This EAL applies to potential emergency conditions which might develop during use of the Independent Spent Fuel Storage Installation and dry cask storage system. This EAL provides for an Unusual Event classification, which may be entered in the event that conditions occur which have the potential for damaging or degrading the fuel, but no releases of radioactive material requiring offsite response or monitoring are expected. Consistent with the NUMARC guidance, escalations above the Unusual Event are not warranted.

Accidents associated with the dry cask storage system include natural and man-made events that are postulated to affect the storage system. The limiting impacts to the system include loss of shielding capability and loss of confinement. The loss of shielding results in higher direct radiation to the environment from the cask while the loss of confinement results in a release of materials from within the cask to the environment at a postulated leak rate.

Loss of confinement for the dry storage system is evaluated in TN-68, Safety Analysis Report, Section 7. Two scenarios are considered, one for off-normal conditions and one for hypothetical accident conditions. Dose calculations are included in section 7.3.2.1. In the extremely unlikely event that one of these scenarios did occur, the event would be addressed by the Radioactivity Release EALs contained in Table 5.

Loss of shielding for the dry storage system is evaluated in TN-68, Safety Analysis Report, Section 5. Dose calculations are included in Table 5.1-2 for both normal and accident conditions. The value of 600 mR/hr one foot away OR 1200 mR/hr at the external surface are determined for several reasons. According to the TN-68, Safety Analysis Report, Table 5.1-2, Summary of Average Dose Rates, the maximum expected surface dose rates will be 529.5 mR/hr (see note 2). Consequently, the value of 1200 mR/hr is sufficiently above normal conditions as to preclude inappropriate classifications.

Also, the value of 1200 mR/hr is sufficiently below the 1467 mR/hr found in Table 5.1-2 for the cask surface radiological reading for accident conditions. Therefore, 1200 mR/hr from a loss of shielding accident would trigger an Unusual Event classification.

ERP-101 BASES, Rev 3 Page 15 of 133 DEVIATION None REFERENCES NUMARC NESP-007, AU2.3

ERP-101 BASES, Rev 3 Page 16 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.a IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1)

MODE All BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2, "Unexpected Rise in Plant Radiation or Airborne Concentration."

NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for this EAL. The areas where Irradiated fuel is located forms the basis for the radiation monitors listed in Table 1-1.

Unexpected radiation levels which are at least 100 times higher than the normal background will generally indicate a fuel handling accident or loss of water covering the irradiated fuel.

Readings may be from refuel floor Area Radiation Monitors or taken during a qualified radiological survey. Table 1-1 monitors are as follows:

Table 1-1 Refuel Floor ARMs 3-7 (7-9) Steam Separator Pool 3-8 (7-10) Refuel Slot 3-9 (7-11) Fuel Pool 3-10 (7-12) Refueling Bridge There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

ERP-101 BASES, Rev 3 Page 17 of 133 Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel-handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor. Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore appropriate. This radiation level could also be caused by an inadvertent criticality and is included even though the probability of this event occurring is low. Radiation increases above 500 mR/hr which were expected should not cause an Alert to be declared during a planned evolution. Additionally, surveys which identify "hot spots" greater than 500 mR/hr should not cause an Alert to be declared.

Escalation, if appropriate, would occur via Effluent Release, In-plant radiation, or Emergency Director Judgment.

DEVIATION None REFERENCES NUMARC NESP-007, AA2.1 NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents

ERP-101 BASES, Rev 3 Page 18 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.b IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL F=Report of visual observation of irradiated fuel uncovered MODE All BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2, "Unexpected Rise in Plant Radiation or Airborne Concentration."

NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for this EAL.

Studies of the loss of fuel pool water level indicate that a significant release may occur if rapid oxidation of the fuel clad occurs due to prolonged fuel uncovery. Offsite doses are not; however, expected to exceed EPA PAGs. In addition, NRC Information Notice No. 90-08, "Kr 85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel-handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor. Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, In-plant radiation, or Emergency Director Judgment.

ERP-101 BASES, Rev 3 Page 19 of 133 DEVIATION None REFERENCES NUMARC NESP-007, AA2.2

ERP-101 BASES, Rev 3 Page 20 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.c IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL Water Level < 458 "above RPV instrumentzero for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering MODE 5 (With Reactor Refueling Cavity Flooded)

BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2, "Unexpected Rise in Plant Radiation or Airborne Concentration."

NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for this EAL.

There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel-handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor. Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

The value 458" above RPV instrument zero is the Tech. Spec. Limit and an uncontrolled level decrease that would uncover irradiated fuel is an indicator of a decrease in the level of safety of the plant.

ERP-101 BASES, Rev 3 Page 21 of 133 Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, In-plant radiation, or Emergency Director Judgment.

DEVIATION The MODE applicability [5 With Reactor Refueling Cavity Flooded] is a deviation from NUMARC [all] in that the EAL is only applicable in that plant condition. This adds clarity to the EAL to ensure that it will not be applied under plant conditions where a classification is not warranted.

REFERENCES NUMARC NESP-007, AA2.3 Tech Spec 3.9.6

ERP-101 BASES, Rev 3 Page 22 of 133 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.d IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL Water Level < 232 ft 3 inches plant elevation for the Spent Fuel Pool that will result in Irradiated Fuel uncovering MODE All BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2, "Unexpected Rise in Plant Radiation or Airborne Concentration."

NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for this EAL.

There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel-handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor. Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

The value 232 ft 3 inches plant elevation is the Tech. Spec. Limit and an uncontrolled level decrease that would uncover irradiated fuel is an indicator of a decrease in the level of safety of the plant.

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, In-plant radiation, or Emergency Director Judgment.

ERP-101 BASES, Rev 3 Page 23 of 133 DEVIATION None REFERENCES NUMARC NESP-007, AA2.4 Tech Spec 3.7.7

ERP-101 BASES, Rev 3 Page 24 of 133 2.0 Reactor Pressure Vessel 2.1 Reactor Pressure Boundary UNUSUAL EVENT - 2.1.1 IC Reactor Coolant System Leakage EAL The following conditions exist:

Unidentified Primary System Leakage > 10 gpm into the Drywell OR Identified Primary System Leakage > 25 gpm into the Drywell MODE 1,2,3,4 BASIS Utilizing the leak before break methodology, it is anticipated that there will be indication(s) of minor reactor coolant system boundary integrity loss prior to this fault escalating to a major leak or rupture. Detection of low levels of leakage while pressurized is utilized to monitor for the potential of catastrophic failures. Leakage not associated with catastrophic failure potential such as SRV leakage, should not be considered in this EAL.

Identified and unidentified Primary System Leakage is measured by the normal primary system leakage monitoring system and is leakage into the drywell.

This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, it is considered to be a potential degradation of the level of safety of the plant. The value of 10 gpm unidentified leakage is significantly higher than the expected pressurized leak rate from the reactor coolant system. The 10 gpm value for the unidentified pressure boundary leakage was selected as it is twice the Technical Specification value, indicating an increase beyond that assumed in Safety Analysis. It also is observable with normal control room indications. The EAL for identified leakage is set at a higher value (25 gpm) due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.

Technical Specification LCO required actions would necessitate a plant shutdown and subsequent depressurization, unless the source of the leak can be isolated, identified, and/or stopped. Actions initiated by plant staff would include close monitoring of the calculated break size such that any sudden or gradual increase in leak rate would be identified. A slow power reduction and gradual depressurization would be necessitated due to the possibility that a sudden power and/or pressure surge could potentially worsen the break or cause a catastrophic failure.

The leak rate of 10 gpm may cause a high drywell pressure indication. Other indications of a leak of this magnitude would include an increase in drywell temperature or radiation.

ERP-101 BASES, Rev 3 Page 25 of 133 This event will escalate to an Alert based upon high Drywell pressure per Fission Product Barrier Table.

DEVIATION NUMARC Example EAL SU5.1.a identifies pressure boundary leakage. There is no Peach Bottom EAL listed for pressure boundary leakage specifically since it is a subset of unidentified leakage. Peach Bottom Tech. Specs. requires a shutdown if any pressure boundary leakage is found.

REFERENCES NUMARC NESP-007, SU5 Technical Specifications 3.6.C.1 T-101, RPV Control T-102, Primary Containment Control

ERP-101 BASES, Rev 3 Page 26 of 133 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level SITE AREA EMERGENCY - 2.1.3 IC Loss of Water Level in the Reactor Vessel That Has or Will Uncover fuel in the Reactor Vessel EAL RPV level <-172" MODE 4, 5 BASIS The indicator for "core is or will be uncovered" is Reactor Pressure Vessel Water level below the Top of Active Fuel (TAF) -172 inches as indicated on RPV Fuel Zone Level Instruments LI 2(3)-02-3-091 or LI-2(3)-02-3-113. Core submergence ensures adequate core cooling. When RPV level decreases below the top of active fuel the ability to remove the decay heat generated from the nuclear fuel becomes suspect and the Fuel Clad Fission Product barrier can no longer be considered intact. Sustained partial or total core uncovery can result in the release of a significant amount of fission products to the reactor coolant.

Under the conditions specified by this IC, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured. It is intended to address concerns raised by NRC Office for Analysis and Evaluation of Operational Data (AEOD) report AEOD/EG09, "BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel," dated August 8, 1986. This report states:

In broadest terms, the dominant causes of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting form the shutdown cooling mode.

During this transitional period, water is drawn from the reactor vessel, cooled by the residual heat removal system heat exchangers (from the cooling provided by the service water system), and returned to the reactor vessel. First, there are piping and valves in the residual heat removal system which are common to both the shutdown cooling mode and other modes of operation such as low pressure coolant injection and suppression pool cooling. These valves, when improperly positioned, provide a drain path for reactor coolant to flow from the reactor vessel to the suppression pool or the radwaste system. Second, establishing or making such evolutions vulnerable to personnel and procedural errors. Third, there is no comprehensive valve interlock arrangement for all shutdown cooling. Collectively, these factors have contributed to the inadvertent draining of the reactor vessel.

Thus, declaration of a Site Area Emergency is warranted under the conditions specified by the IC. Escalation to a General Emergency is via effluent release EAL.

ERP-101 BASES, Rev 3 Page 27 of 133 DEVIATION During EAL review and approval process, it was determined that the condition stated in NUMARC NESP-007, SS5, l.a "Loss of all decay heat removal cooling as determined by (site specific) procedure" is not necessary to conclude that the plant condition warrants a Site Area Emergency. Therefore, that sample NUMARC EAL was not included in this EAL.

REFERENCES NUMARC NESP-007, SS5

ERP-101 BASES, Rev 3 Page 28 of 133 2.0 Reactor Pressure Vessel 2.2 Reactor Power ALERT - 2.2.2 IC Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EAL Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS SCRAM to make Reactor shutdown MODE 1, 2 BASIS This EAL is not applicable if a manual scram is initiated and no RPS setpoints are exceeded.

Taking the mode switch to shutdown is considered a manual scram action. This may cause an RPS Setpoint to be exceeded due to the change in Nuclear Instrumentation Scram setpoint when the mode switch is placed in shutdown. If the RPS then fails to initiate a scram, then this should be evaluated as an automatic RPS setpoint being exceeded.

Entry into this EAL is based on a reactor parameter actually exceeding a RPS setpoint and the reactor is not brought to a sub-critical condition and maintained at that state with automatic RPS functions. The parameter must exceed the RPS setpoint by a significant margin eliminating minor setpoint drifts which are accounted for in the Technical Specification Margin of Safety. Subsequent manual scram actions were successful in bringing the reactor to a shutdown condition. Confirmation indications include control room annunciators, APRM/WRPM power level, and Control rod position indication.

A failure of the Reactor Protection System (RPS) to initiate and complete a reactor scram may indicate that the design limits of the nuclear fuel has been compromised. RPS is designed to automatically detect and generate a reactor scram signal when a limiting safety system setpoint is reached or exceeded. Control rod insertion following a scram signal is designed to be passive (i.e., system de-energizes, control rod motive energy source is previously charged).

Assuming that shutdown (sub-critical) conditions cannot be established/maintained, an automatic scram signal failure followed by a successful manual scram would still constitute a scram failure and should be classified under this event.

Although the reactor may be brought initially subcritical based on partial control rod insertion, there is a possibility that positive reactivity may be introduced by a number of factors. Xenon decay and factors associated with cooldown, lower fuel temperature (doppler), lower moderator temperature, and a lower presence of steam bubbles (voids) may all contribute to cause a power increase.

ERP-101 BASES, Rev 3 Page 29 of 133 Subcritical conditions can be assured even with the most reactive control rod fully withdrawn from the core if the remaining 184 control rods fully insert. Any other control rod pattern resulting from partial control rod insertion must be carefully analyzed and/or monitored to detect the possibility of re-criticality or local criticality.

Due to the buildup of Xenon in areas of the core that have previously been operating at high power levels, attention should be applied to the possibility that control rods which previously had low worth (e.g., peripheral control rods) may now have significant control rod worth.

When the reactor is not shutdown as identified in the Transient Response Implementing Plan Procedures (TRIPs), then entry into this EAL is warranted. When partial control rod insertion occurs following a scram signal (either manual or automatic) judgment should be applied as to whether classification should occur. Multiple control rods failing to insert beyond notch position 02 may require actions to fully insert the control rods. However, the reactor has been made subcritical, and for all intent the reactor will remain subcritical. TRIP guidance will govern the insertion of these control rods.

This EAL would be escalated with a failure of both manual and automatic scram signals with the Reactor remaining critical.

DEVIATION None REFERENCES NUMARC NESP-007, SA2 T-101, RPV Control, RC-1

ERP-101 BASES, Rev 3 Page 30 of 133 2.0 Reactor Pressure Vessel 2.2 Reactor Power SITE AREA EMERGENCY - 2.2.3 IC Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EAL RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

MODE 1,2 BASIS This EAL is not applicable if a manual scram is initiated and no RPS setpoints are exceeded.

Taking the mode switch to shutdown is considered a manual scram action. This may cause an RPS Setpoint to be exceeded due to the change in Nuclear Instrumentation Scram setpoint when the mode switch is placed in shutdown. If the RPS then fails to initiate a scram, then this should be evaluated as an automatic RPS setpoint being exceeded.

A valid automatic and/or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 4% power. The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is "fail safe," that is, it de-energizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure).

A failure of the Reactor Protection System to shut down the reactor (as indicated by reactor power remaining above 4%) is a degraded plant condition that together with suppression pool temperature approaching 1 10°F requires the injection of boron to shut down the reactor.

The RPV Control TRIP Procedure establishes 4% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. With Reactor Power less than 4% the heat being generated in the core can be removed from the RPV and containment while actions are taken to bring the reactor subcritical.

A manual scram is defined as any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical (i.e., mode switch to shutdown, manual scram push buttons, or manual ARI initiation). Taking the mode switch to shutdown as part of the actions required by TRIP procedure is considered a manual scram action.

ERP-101 BASES, Rev 3 Page 31 of 133 While the plant is being shutdown, significant heat is being generated in the core and the heat up rate of the Torus (due to heat rejection through SRVs) can increase which could approach the Torus temperature limit prior to shutting down. As the Torus heat increases towards the limiting temperature, the probability of causing a major over-pressure event increases substantially.

After an ATWS event, there is a potential that the Main Steam Isolation Valves (MSIV) will remain open. There is additional guidance in the TRIP procedures to ensure that the MSIVs remain open even if RPV level is intentionally lowered to below the normal MSIV isolation level.

This situation would allow the plant to remove heat and provide makeup through the normal steam/feed cycle. If this path is not available, or becomes unavailable during the transient, heat rejection will be to the Torus.

With Standby Liquid Control initiated and with partial or no control rod insertion, there is a possibility that the neutron flux profile in the reactor core may become uneven or distorted.

Localized clad damage is possible, if localized power levels increase significantly.

With reactor power remaining above 4% containment integrity is threatened, as the ability of systems to remove all of the heat transferred to the containment may be exceeded. As the energy contained in the containment increases there may be a degradation in the ability to remove heat generated by the "at power" reactor core. There is therefore a potential loss of the containment or the fuel cladding (caused by overheating).

This event will be escalated based on Torus Temperature on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-102, T/T-1) or RPV level <-200".

DEVIATION None REFERENCES NUMARC NESP-007, SS2 T-100, Scram T-101, RPV Control, RC/L-2 T-1 17, Level/Power Control

ERP-101 BASES, Rev 3 Page 32 of 133 2.0 Reactor Pressure Vessel 2.2 Reactor Power GENERAL EMERGENCY - 2.2.4 IC Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EAL RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

AND Torus Temperature is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-102, T/T-1) OR RPV level <-200" MODE 1, 2 BASIS Taking the mode switch to shutdown is considered a manual scram action. This may cause an RPS Setpoint to be exceeded due to the change in Nuclear Instrumentation Scram setpoint when the mode switch is placed in shutdown. If the RPS then fails to initiate a scram, then this should be evaluated as an automatic RPS setpoint being exceeded.

When Torus level is outside of the Heat Capacity Temperature Limit Curve High or Low, it is appropriate to consider operation to be on the "UNSAFE" side.

A valid automatic or manual scram signal is present as indicated by control room indications and/or alarms and APRM indication is greater than 4% power. In addition, control room instrumentation indicates that operation is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-102, T/T-1) or RPV level is < -200".

Failure of all automatic and manual trip functions coincident with a high Torus temperature will place the plant in a condition where reactivity control capability is jeopardized and heat removal capability is severely limited.

RPV level <-200 " indicates an extreme challenge to the ability to cool the core.

The RPV Control TRIP Procedure establishes 4% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. The timely initiation of Standby Liquid Control (prior to Torus temperature reaching 11 0°F) would bring the reactor to < 4% power before Torus temperature approaches the heat capacity temperature limit curve limitations.

ERP-101 BASES, Rev 3 Page 33 of 133 Under ATWS conditions, it is important to assure continuous, stable steam condensation capability. An elevated Torus temperature on the "UNSAFE" side of the HCTL curve would result in unstable steam condensation should rapid reactor depressurization occur (ADS activation). Maintaining the ability to condense steam will preclude the pressurization of the containment and prevent possible containment failure Containment over-pressurization, which would be an eventual result of sustained operation with heat being added to the containment and high Torus temperature would result in the loss of containment integrity and the inability to remove the heat generated from the fuel. Fuel clad failure would result from the overheating of the fuel.

DEVIATION None REFERENCES NUMARC NESP-007, SG2.1, SG2.2 T-101, RPV Control T-102, Primary Containment Control, T/T-1 T-1 17, Level/Power Control, RC/L-2

ERP-101 BASES, Rev 3 Page 34 of 133 3.0 Fission Product Barrier 3.1 Initiating Condition Matrix Determine which combination of the three barriers (Fuel Clad, Reactor Coolant, Primary Containment) are lost or have a potential loss and use the following key to classify the event.

Also, an event for multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds is IMMINENT (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this IMMINENT LOSS situation, use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT IC ANY Loss or ANY Potential Loss of Primary Containment EAL ANY Loss OR ANY Potential Loss of Primary Containment ALERT IC ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS EAL ANY Loss OR ANY Potential Loss of EITHER Fuel Clad OR RCS SITE AREA EMERGENCY IC Loss of BOTH Fuel Clad AND RCS OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS, and Loss of ANY Additional Barrier EAL Loss of BOTH Fuel Clad AND RCS OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS, AND Loss of ANY Additional Barrier

ERP-101 BASES, Rev 3 Page 35 of 133 GENERAL EMERGENCY IC Loss of ANY Two Barriers AND Potential Loss of Third Barrier EAL Loss of ANY Two Barriers AND Potential Loss of Third Barrier MODE 1, 2,3 NOTES:

1. Although the logic used for these initiating conditions appears overly complex, it is necessary to reflect the following considerations:

The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Containment barrier. Unusual Event ICs associated with RCS and Fuel Clad barriers are addressed under the other plant condition EALs.

At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from General Emergency. For example, if the Fuel Clad barrier and RCS barrier "Loss" EALs existed, this would indicate to the Emergency Director that, in addition to offsite dose assessments, must focus on continual assessments of radioactive inventory and containment integrity. If, on the other hand, both Fuel Clad barrier and RCS barrier "Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

The ability to escalate to higher emergency classes as an event gets worse must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.

2. Fission Product Barrier ICs must be capable of addressing event dynamics. Thus, the EAL Reference Table states that IMMINENT (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) Loss or Potential Loss should result in a classification as if the affected threshold(s) are already exceeded, particularly for the higher emergency classes.

3. The Fuel Clad barrier is the cladding tubes that contain the fuel pellets.

4. The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

5. The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.

ERP-101 BASES, Rev 3 Page 36 of 133

6. If a "Loss" condition is satisfied, the "Potential Loss" category can be considered satisfied. This is also applicable to conditions where this is a "Loss" indication with no corresponding "Potential Loss" condition.

7. For all conditions listed in Fission Product Barrier Table, the barrier failure column is only satisfied if it fails when called upon to mitigate an accident. For example, failure of both containment isolation valves to isolate with a downstream pathway to the environment is only a concern during an accident. If this condition exists during normal power operations, it will be an active Technical Specification Action Statement.

However, during accident conditions, this will represent a breach of containment.

DEVIATION None REFERENCES NUMARC NESP-007, Recognition Category F, Table 3

ERP-101 BASES, Rev 3 Page 37 of 133 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.1 Primary Coolant Activity Level EAL LOSS Reactor Coolant activity > 300 ,Ci/gm Dose Equivalent Iodine 131 POTENTIAL LOSS Not Applicable MODE 1, 2,3 BASIS A reactor coolant sample activity of greater than > 300 .iCi/gm was determined to indicate significant clad heating and is indicative of the loss of the fuel clad barrier. This concentration is well above that expected for Iodine spikes and corresponds to 2.6% clad damage. 2.6%

fuel clad damage is based upon NUREG-1228 core damage analysis.

Calculation of 300 I+/-Ci/cc equivalence to percent fuel clad damage is as follows (for purposes of this calculation, cc and gm are considered equivalent):

Iodine Isotope Dose Factors Ci/MWe Values (Time After Shutdown = 0)

(Req Guide 1.109) (NUREG-1228) 1-131 4.39E-3 85000 1-132 5.23E-5 120000 1-133 1.04E-3 170000 1-134 1.37E-5 190000 1-135 2.14E-4 150000 Time After Shutdown (T = 0) Ratios R132 = 120000/85000(1-131) = 1.41 (1-131)

R133 = 170000/85000(1-131) = 2.00(1-131)

R134 = 190000/85000(1-131) = 2.24(1-131)

R135 = 150000/85000(1-131) = 1.76(1-131)

Equation for Dose Equivalent Iodine (DEl131)

A in DF 131 + (R 132) A 13) DF 12 + (R 133) A DF 133 M31 + (R 134) A 131 DF134 + (R 13s) A 131 DE1131 DF 131

ERP-101 BASES, Rev 3 Page 38 of 133 Solve for A13 1 assuming DEl 131 = 300 jtCi/cc A 1314. 39E - 3+1.41 A 1315. 23E -5+2.00 A 131. 04E - 3+2.24 A 1311.37E 1.76 A 13j2.14E 300 =

4.39E - 3 300 = 6.95E - 3 A13 4.39E -3 Therefore: A131 = 189 p.Ci/cc 1-131 Clad damage fraction (NUREG-1228, Table 4.1) = .02 Full Power = 1150 MWe Clad Activity 1-131 = (Ci/MWe) (MWe) (Clad Damage Fraction)

= (85000Ci/MWe) (1 150MWe) (.02)

= 1.96E6 Ci Reactor Water Volume = 2.67E8 cc (ERP-C-1410)

Total Coolant Activity 1-131 = (A131) (Rx Water Volume) (Ci/R.Ci)

= (189 RiCi/cc) (2.67E8cc) (1.OE-6Ci/ [tCi)

= 5.05E4Ci Percent Clad Damage = Total Coolant Activity/Clad Activity 1-131

= (5.05E4) / (1.96E6)

= 2.6%

This event will be escalated to an Site Area Emergency when additional fission product barriers are lost.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #1 NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, Table 2.2 Reg. Guide 1.109, Table E-9 ERP-C-1410

ERP-101 BASES, Rev 3 Page 39 of 133 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.2 Reactor Vessel Water Level EAL LOSS RPV level < -200" POTENTIAL LOSS RPV level < -172 "

MODE 1, 2,3 BASIS The "Loss" EAL -200 " value corresponds to the level which is used in the TRIPS to indicate challenge of core cooling. This is the minimum value to assure core cooling without further degradation of the clad. The "Potential Loss" EAL is the same as the RCS barrier "Loss" EAL 4 and corresponds to the fuel zone water level at the top of the active fuel. Thus, this EAL indicates a "Loss" of RCS barrier and a "Potential Loss" of the Fuel Clad Barrier. This EAL appropriately escalates the emergency class to a Site Area Emergency.

Core submergence is the preferred method of core cooling and as such, the failure to re establish RPV water level above the top of active fuel for an extended period of time could lead to significant fuel damage. This condition, -200 " as read on instruments LI-2(3)-02-3-091 or L12(3)-02-3-113, could be indicative of a large break Loss Of Coolant Accident (LOCA)

(where ECCS Systems are designed to maintain level at 2/3 core height) or a small LOCA with the inability of emergency core cooling systems to reflood the RPV. The value of -200" was chosen as it represents 2/3 core height.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 T-101, RPV Control T-1 11, Level Restoration/Steam Cooling, LR-1 1 "T-112, Rapid Depressurization T-1 17, Level/Power Control T-116, RPV Flooding

ERP-101 BASES, Rev 3 Page 40 of 133 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.3 Drywell Radiation Monitoring EAL LOSS Drywell Rad Monitor reading > 8x10 4 R/hr POTENTIAL LOSS Not Applicable MODE 1, 2,3 BASIS The 8x10 4 R/hr reading on a containment high range radiation monitor RI-8(9)103A,B,C,D is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading was calculated assuming an instantaneous release and dispersal of the Reactor Coolant noble gas and iodine inventory into the Primary Containment (direct reading not shine) at a coolant concentration of 300 jICi/gm Dose Equivalent Iodine 131. This calculation is as follows:

Using Curve 3 [1%] of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 30,000 R/hr Extrapolating to 2.6%

(30,000 R/hr/1%)(2.6) = 78,000 R/hr This is rounded conservatively to 80,000 R/hr for human factors considerations 2.6% clad damage is based upon NUREG-1228 core damage analysis, and by virtue of its release into containment, the loss of the Reactor Coolant barrier (detailed calculations are contained in the Basis for Fission Product Barrier EAL FC #1).

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss EAL #3. Thus, this EAL indicates a loss of both Fuel Clad barrier and RCS barrier.

There is no "Potential Loss" EAL associated with this item.

ERP-101 BASES, Rev 3 Page 41 of 133 DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #3 and RC EAL #3 NUREG 1228 - Source Term Estimation During Incident Response to Nuclear Power Plant Accidents ERP-C-1410

ERP-101 BASES, Rev 3 Page 42 of 133 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.4 Other Indications EAL LOSS Not Applicable POTENTIAL LOSS Not Applicable MODE 1, 2,3 BASIS There are no other indications at PBAPS for loss of the Fuel Clad Barrier.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4 and RC EAL #5

ERP-101 BASES, Rev 3 Page 43 of 133 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.5 Emergency Director Judgment EAL Any condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the FUEL CLAD barrier MODE 1, 2, 3 BASIS This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL, as a factor in Emergency Director judgment, that the barrier may be considered lost or potentially lost. (See also IC, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #5

ERP-101 BASES, Rev 3 Page 44 of 133 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.1 RCS Leak Rate EAL LOSS Not Applicable POTENTIAL LOSS RCS leakage >50 gpm OR Unisolable primary system leakage outside drywell as indicated by T-103, Temperature Action Level is exceeded in ONE area requiring a SCRAM OR Unisolable primary system leakage outside drywell as indicated by T-103, Radiation Action Level is exceeded in ONE area requiring a SCRAM MODE 1, 2,3 BASIS UNISOLABLE - A leak that cannot be isolated from the Control Room.

When evaluating this EAL for unisolable primary system leakage, it is appropriate to attempt isolation from the Control Room prior to classification.

The potential loss of RCS based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak, however, a break propagation leading to a significantly larger loss of inventory is possible. RCS leakage is measured by the normal primary system leakage monitoring system and is leakage into the drywell. Under certain conditions, this system may be isolated due to increased drywell pressure caused by the leak. In that case, a "loss" of RCS will be indicated and this "potential loss" of RCS would not impact the classification.

Inventory loss events, such as a stuck open SRV, should not be considered when referring to "RCS leakage" because they are not indications of a break which could propagate.

Potential loss of RCS based on primary system leakage outside the drywell is determined from T-103 area temperatures or radiation levels. TRIP guidance stipulates that when the Temperature or Radiation Action Level limits have been exceeded for one area, that the reactor be manually scrammed.

ERP-101 BASES, Rev 3 Page 45 of 133 There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e., primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the Temperature Action Levels Isolation levels. If the temperatures continue to rise to the Temperature Action Levels it is indicative that an unisolable leak has occurred. If the radiation levels rise above the Radiation Action Levels, it also indicates that an unisolable leak has occurred.

This event signifies that there is a direct path established for the transfer of main steam to inside the Turbine Building. Assumptions made in dose calculations regarding radioactive material transport (e.g., hold up, plate out, scrubbing, and retention) may be invalid.

Additionally the transport time associated with a radiological release may be significantly shortened and there may be a higher percentage of short lived radioisotopes in any release.

As both the reactor coolant pressure boundary and the primary containment are degraded; the extent of radioactive release is dependent on fuel clad integrity. Should a rapid reactor depressurization occur as a result of this event then there is a potential that a large amount of radioiodine may be released.

DEVIATION None REFERENCES NUMARC NESP-007, RC EAL #1 PC EAL #2 T-103 Secondary Containment Control

ERP-101 BASES, Rev 3 Page 46 of 133 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.2 Drywell Pressure EAL LOSS Drywell Pressure > 2.0 psig AND Indication of a leak inside drywell POTENTIAL LOSS Not Applicable MODE 1, 2, 3 BASIS The 2.0 psig drywell pressure is based on the drywell high pressure alarm set point and indicates a LOCA.

If drywell pressure exceeds 2 psig, there is a clear indication that a leak of sufficient magnitude exists that prevents drywell pressure stabilization.

DEVIATION The NUMARC EAL contains only the drywell pressure. A qualifying:

"AND Indication of a leak inside drywell" was added as a human factor reminder to the Emergency Director that use of this EAL is for accident scenarios only. Thus, a Drywell pressure increase due to the loss of Drywell cooling will not require an emergency classification.

REFERENCES NUMARC NESP-007, RC EAL#2 T-101, RPV Control T-102, Primary Containment Control

ERP-101 BASES, Rev 3 Page 47 of 133 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.3 Drywell Radiation Monitoring EAL LOSS Drywell Rad Monitor reading > 15 R.1hr POTNIAL LOSS Not Applicable MODE 1, 2,3 BASIS The 15 R/hr reading is a value which indicates the release of reactor coolant to the drywell.

The value assumes an instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with concentrations corresponding to 0.001% Total Isotopic Distribution (TID) into the drywell atmosphere.

Using attachment 5 of ERP-C-1410, Curve 6 Time after Shutdown = 0.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 0.001% TID = 17 R/hr This is rounded to 15 R/hr for human factors considerations This reading is less than that specified for Fuel Clad Barrier EAL #3. Thus, this EAL would be indicative of a RCS leak only. If the radiation monitor reading increases to that value specified by Fuel Clad Barrier EAL #3, fuel damage would also be indicated.

There is no "Potential Loss" EAL associated with this item.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #3 and RC EAL #3 NUREG 1228 - Source Term Estimation During Incident Response to Nuclear Power Plant Accidents ERP-C-1410, Attachment 5

ERP-101 BASES, Rev 3 Page 48 of 133 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.4 Reactor Vessel Water Level EAL LOSS RPV level < -172" POTENTIAL LOSS Not Applicable MODE 1, 2, 3 BASIS This "Loss" EAL is the same as "Potential Loss" Fuel Clad Barrier EAL #2. The -172 " water level corresponds to the level which is used in TRIPS to indicate challenge of core cooling.

This EAL appropriately escalates the emergency class to a Site Area Emergency. Thus, this EAL indicates a loss of the RCS barrier and a Potential Loss of the Fuel Clad Barrier.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 T-101, RPV Control T-1 11, Level Restoration/Steam Cooling, LR-1 1 T-112, Rapid Depressurization T-1 17, Level/Power Control T-116, RPV Flooding

ERP-101 BASES, Rev 3 Page 49 of 133 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.5 Other Indications EAL LOSS Not Applicable POTENTIAL LOSS RPV level cannot be determined MODE 1, 2, 3 BASIS Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled parameter oscillations.

TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design, all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV level indication exists. Multiple indications of level instruments pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.

If indeterminate RPV level is due to reference leg boil-off, it is an indicator of a potential loss of the Reactor Coolant System. Adequate core cooling would be rapidly assured using the guidance provided in the TRIP Procedures. If it can be determined that the cause is due to an instrument power or instrumentation failure, then it is not appropriate to classify the event as a potential loss of the Reactor Coolant System.

Operator attention should be given to the possibility that under depressurized conditions, there is the possibility that gases may come out of solution and result in distorted RPV level indications. Operators should be attentive to observe multiple level indications (particularly those which utilize separate reference legs) to ensure that actual RPV level is known and displayed. Unexplained and/or sudden changes in specific level indications may be a result of degassification of the coolant contained in the level instrumentation.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4 and RC EAL #5 T-101, RPV Control, RC/L-1 T-112, Rapid Depressurization T-1 17, Level/Power Control T-116, RPV Flooding

ERP-101 BASES, Rev 3 Page 50 of 133 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.6 Emergency Director Judgment EAL Any condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the RCS barrier MODE 1, 2,3 BASIS This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost. (See also IC, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

DEVIATION None REFERENCES NUMARC NESP-007, RCS EAL #6

ERP-101 BASES, Rev 3 Page 51 of 133 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.1 Drywell Pressure EAL LOSS Rapid, unexplained drop in Drywell Pressure following initial rise OR Drywell pressure response not consistent with LOCA conditions POTENTIAL LOSS Drywell Pressure > 49 psig and rising OR Drywell Hydrogen > 6% AND Drywell Oxygen > 5%

MODE 1, 2,3 BASIS Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity- Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity. The 49 psig for potential loss of containment is based on the containment drywell design pressure and is equal to the peak pressure expected from a DBA LOCA.

The specified value of 6% hydrogen concentration is the minimum which can support a deflagration. Likewise, the minimum concentration of oxygen required to support a deflagration is 5%. Combustion of hydrogen in the deflagration concentration range creates a traveling flame causing a rapid rise in primary containment pressure. A deflagration may result in a peak primary containment pressure high enough to rupture the primary containment or damage the drywell-to-torus boundary.

DEVIATION None REFERENCES NUMARC NESP-007, PC EAL #1 ON-110, Loss of Primary Containment T-101, RPV Control T-102, Primary Containment Control w/Bases T-1 03, Secondary Containment Control

ERP-101 BASES, Rev 3 Page 52 of 133 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.2 Containment Isolation Valve After Containment Isolation EAL LOSS Failure of both valves in any one line to close AND downstream pathway to the environment exists OR Intentional venting per T-200 is required OR Unisolable primary system leakage outside drywell as indicated by T-103, Temperature Action Level is exceeded in ONE area requiring a SCRAM OR Unisolable primary system leakage outside drywell as indicated by T-103, Radiation Action Level is exceeded in ONE area requiring a SCRAM POTENTIAL LOSS Not Applicable MODE 1, 2, 3 BASIS UNISOLABLE - A leak that cannot be isolated from the Control Room.

When evaluating this EAL for unisolable primary system leakage, it is appropriate to attempt isolation from the Control Room prior to classification.

This EAL is intended to cover containment isolation failures allowing a direct flow path to the environment such as failure of both MSIVs to close with open valves downstream to the turbine or to the condenser. In addition, the presence of area radiation or temperature alarms indicating unisolable primary system leakage outside the drywell are covered. Also, an intentional venting of primary containment per TRIPS to the secondary containment and/or the environment is considered a loss of containment.

Loss of containment based on primary system leakage outside the drywell is determined from T-103 area temperatures or radiation levels. TRIP guidance stipulates that when the Temperature or Radiation Action Level limits have been exceeded for one area, that the reactor be manually SCRAMmed.

ERP-101 BASES, Rev 3 Page 53 of 133 There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e., primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the Temperature Action Level Isolation levels. If the temperatures continue to rise to the Temperature Action Levels it is indicative that an unisolable leak has occurred. If the radiation levels rise above the Radiation Action Levels, it also indicates that an unisolable leak has occurred.

DEVIATION None REFERENCES NUMARC NESP-007, RCS EAL #1, PC EAL #2 T-103 Secondary Containment Control T-104, Radioactivity Release Control T-200, Primary Containment Venting

ERP-101 BASES, Rev 3 Page 54 of 133 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.3 Significant Radioactive Inventory in Containment EAL LOSS Not Applicable POTENTIAL LOSS Drywell Rad Monitor reading > 6x10 5 Rlhr MODE 1, 2,3 BASIS A containment high range radiation monitor 9RI-8(9)103A,B,C,D reading 6x10 5 R/hr indicates significant fuel damage, well in excess of that required for the loss of the RCS and Fuel Clad.

As stated in Section 3.8 of NUMARC/NESP-007, a major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents,"

indicates that such conditions do not exist when the amount of clad damage is less than 20%.

The reading was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment (direct reading not shine) where the value corresponds to a release of approximately 20% of the gap region. This calculation is as follows:

Using Curve 3 [1%1 of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 30,000 R/hr Extrapolating to 20%

(30,000 R/hr/1%)(20) = 600,000 R/hr There is no "Loss" EAL associated with this item.

ERP-101 BASES, Rev 3 Page 55 of 133 DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #3, RC EAL #3 and PC EAL #3 NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents ERP-C-1410

ERP-101 BASES, Rev 3 Page 56 of 133 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.4 Reactor Vessel Water Level EAL LOSS Not Applicable POTENTIAL LOSS RPV level cannot be restored above -200 "within the time limit of the "SAFE" region of the Maximum Core Uncovery Time Limit Curve (T-116, RF-1)

MODE 1, 2,3 BASIS In this EAL, the -200 " water level corresponds to the level which is used in the TRIPS to indicate challenge of core cooling. This is the minimum value to assure core cooling without further degradation of the clad.

When evaluating this EAL for a time after shutdown of less than 90 minutes, it is appropriate to use the value of the Maximum Core Uncovery Time Limit Curve at 90 minutes after shutdown.

The conditions in this potential loss EAL represent imminent melt sequences which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with the level EALs in the Fuel and RCS barrier columns, this EAL will result in the declaration of a General Emergency on loss of two barriers and the potential loss of a third. If the TRIPS have been ineffective in restoring reactor vessel level within the maximum core uncovery time limit, there is not a "success" path.

Severe accident analysis (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events.

Given this, it is appropriate to provide a reasonable period to allow TRIPS to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within the time provided by the maximum core uncovery time limit. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be, ineffective.

There is no "Loss" EAL associated with this item.

DEVIATION None

ERP-101 BASES, Rev 3 Page 57 of 133 REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 T-101, RPV Control T-1 11, Level Restoration/Steam Cooling, LR-1 1 T-1 12, Rapid Depressurization T-1 17, Level/Power Control T-1 16, RPV Flooding

ERP-101 BASES, Rev 3 Page 58 of 133 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.5 Other Indications EAL LOSS Not Applicable POTENTIAL LOSS RPV level cannot be determined AND RPV Flooding cannot be established as indicated by inability to maintain 5 ADS/SRVs open with RPV pressure at least 60 psig above Torus pressure per T-1 16 MODE 1, 2,3 BASIS The decision to enter RPV Flooding is made when RPV water level cannot be determined. This judgment consists of evaluating all plant indications which can influence the ability to maintain adequate core cooling. Entry to RPV flooding requires rapid RPV depressurization. The minimum RPV Flooding Pressure is defined as the lowest differential pressure between the RPV and the Torus at which steam flow through the SRVs will be sufficient to remove all of the generated decay heat. Operation at the minimum reactor flooding pressure requires that a sufficient amount of water reach the core to carry away decay heat by boiling, which in turn requires that RPV water level increase. So RPV flooding not established requires containment flooding. This represents a potential loss of containment due to the potential need to vent containment in order to facilitate flooding. Additionally, it represents a potential inability to remove decay heat which may also lead to containment failure.

Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled indication oscillations. TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design, all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV level indication exists. Level indication pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.

If it can be determined that the loss of ability to monitor RPV level is due to an instrument power or instrumentation failure, then it is not appropriate to classify the event as a potential loss of the Primary Containment.

ERP-101 BASES, Rev 3 Page 59 of 133 The minimum RPV flooding pressure will ensure that adequate core cooling exists independent of RPV level indication. Failure to establish the differential pressure between the RPV and the Torus in a timely manor can jeopardize the ability of the reactor coolant system to dissipate the decay heat generated.

Ample time must be allotted for determining the failure of ECCS systems to pressurize the RPV. Control Room indications such as RPV level (used for trending), RPV Pressure, ECCS injection flow rates, Containment parameters, and injection system operability should all be used to gauge the effectiveness of the RPV Flood.

If the loss of level indication was caused by reference leg flashing, then level indicators can still be utilized to monitor the trend in RPV level. Actual RPV level will never be higher than indicated level.

In the event that the loss of level indication is only a result of degassification of the coolant contained in the level instrumentation piping, then it is anticipated that flooding pressure can be obtained.

RPV water level below the top of active fuel for a sustained period of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. For events starting from power operation, some core melting can be expected. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4, RCS EAL #5 and PC EAL #5 T-101, RPV Control T-1 11, Level Restoration/Steam Cooling, LR-1 1 T-112, Rapid Depressurization T-1 17, Level/Power Control T-116, RPV Flooding

ERP-101 BASES, Rev 3 Page 60 of 133 3.0 Fission Product Barrier 3.4 Primary Containment Barrier PC.6 Emergency Director Judgment EAL Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Primary Containment barrier MODE 1, 2,3 BASIS This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Containment Barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost. (See also IC, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

DEVIATION None REFERENCES NUMARC NESP-007, PC EAL #6

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ERP-101 BASES, Rev3 Page 62 of 133 3.6 Fission Product Barrier Status Table 1 -

Barrier Fuel Clad Reactor Coolant System Primary Containment Parameter Reactor Coolant Activity RPV Level RPV Level Unknown RCS Leak Rate Drywell Pressure Drywell Radiation

ERP-101 BASES, Rev 3 Page 63 of 133 3.5 Fission Product Barrier Status Table Barrier Fuel Clad Reactor Coolant System Primary Containment Parameter Potential Loss Loss Los-n-,'.

Containment Unisolable primary system Failure of both valves in any Isolation leakage outside drywell as one line to close AND indicated by T-103, downstream pathway to the Temperature Action Level environment exists is exceeded in ONE area requiring a SCRAM OR OR Intentional venting per Unisolable primary system leakage outside drywell as T-200 is required indicated by T-1 03, OR Radiation Action Level is Unisolable primary system exceeded in ONE area leakage outside drywell as requiring a SCRAM indicated by T-103, Temperature Action Level is exceeded in ONE area requiring a SCRAM OR Unisolable primary system leakage outside drywell as indicated by a T-103, Radiation Action Level is exceeded in ONE area requiring a SCRAM Emergency Director Any condition in the judgment of the Emergency Director Any condition inthe judgment of the Emergency Director Any condition in the judgment of the Emergency Din Judgment that indicates Loss or Potential Loss of the FUEL CLAD that indicates Loss or Potential Loss of the RCS barrier that indicates Loss or Potential Loss of the Primary barrier Containment barrier In the table below, circle all of the appropriate X's in each applicable row for each Loss or Potential Loss of Fission Product Barrier as determined by the table above.

Classify the event as identified in the table heading if all X's in a column under that heading are circled.

Fission Product Barrier Status Unusual ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Fuel Clad - Loss __ X_

X X X X X X X Fuel Clad - Potential Loss X X X X X Reactor Coolant System - Loss X X X X X X X Reactor Coolant System-Potential Loss X X X X X Primary Containment - Loss X X X X X X X X Primary Containment - Potential Loss X X

ERP-101 BASES, Rev 3 Page 64 of 133 4.0 Secondary Containment Bypass 4.1 Main Steam Line UNUSUAL EVENT - 4.1.1 IC Fuel Clad Degradation EAL Main Steam Line HiHi Radiation (1OxNFPB)

MODE 1, 2,3 BASIS Main Steam Line High-High Radiation alarm (2(3)-252,A,B,C,D and 2(3)-251,A,B,C,D) > 10 times normal full power background may be indicative of minor fuel cladding degradation and warrants the declaration of an Unusual Event. This level is set to preclude most spurious events including resin intrusion.

The main steam line high-high radiation condition requires a manual Main Steam Isolation Valve closure and a reactor scram. This transient may result in the introduction of fission product gases (previously contained in the gap area) to be suddenly released into the coolant due to the rapid down power transient and subsequent collapse of voids in the coolant.

This level of steam line activity is indicative of the release of gap activity to the coolant however, this level is not indication of a major failure of the fuel clad. The mechanics that caused main steam line radiation to increase to this level indicate there is a degradation of fuel clad integrity.

This event will escalate to an Alert based on the breach in the main steam line together with a failure of the MSIVs to isolate the main steam lines per Fission Product Barrier Table.

DEVIATION The MODE applicability [1,2,3] is a deviation from NUMARC [all] in that, the SJAE Radiation Monitor and Main Steam Line Radiation Monitors will only be a valid indication of Fuel Clad Degradation in those MODE's. At Peach Bottom, there are no other monitors which can be an indicator of Fuel Clad Degradation. Degradation in cold shutdown or refueling will be first indicated by ventilation release monitors and covered in Effluent Release section.

REFERENCES NUMARC NESP-007, SU4.1 T-099, Post Scram Recovery T-101, RPV Control

ERP-101 BASES, Rev 3 Page 65 of 133 4.0 Secondary Containment Bypass 4.1 Main Steam Line ALERT - 4.1.2 IC RCS Leak Rate EAL Indication of a Main Steam Line Break:

Hi Steam Flow Annunciator AND Hi Steam Tunnel Temperature Annunciator OR Direct report of steam release MODE 1, 2,3 BASIS When evaluating this EAL, "Direct report of steam release" is considered a leak of magnitude and location that is indicative of a Main Steam Line Break.

Design basis accident analyses of a Main Steam Line Break outside of secondary containment shows that even if MSIV closure occurs within design limits, dose consequences offsite from a "puff" release would be in excess of 10 millirem.

Hi Steam Flow Annunciator and Hi Steam Tunnel Temperature Annunciator are both indicators of a Main Steam Line Break. Both parameters will cause an isolation of the MSIV's. Should both valves in any one line fail to isolate, this event would be considered a loss of Primary Containment and a potential loss of the RCS per the Fission Product Barrier Table and appropriately classified as a Site Area Emergency.

Direct report of steam release is meant to provide an alternate means of classification if the Hi Steam Flow Annunciator or the Hi Steam Tunnel Temperature Annunciator fails to operate and the visual observation of conditions indicates a Main Steam Line Break in the judgment of the Emergency Director. This is not meant to cause a declaration based on leaks such as valve packing leaks where the consequences offsite would be negligible.

DEVIATION None REFERENCES NUMARC NESP-007, RC.1 "T-101, RPV Control NUMARC Questions and Answers, June 1993, "Fission Product Barriers #7"

ERP-101 BASES, Rev 3 Page 66 of 133 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT - 5.1.1.a IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer EAL A valid reading on one or more of the following radiation monitors that exceeds TWO TIMES the HiHi alarm setpoint value for > 60 minutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate using computer dose model exceeds 0.114 mRemlhr TPARD OR 0.342 mRemlhr child thyroid CDE based on a 60 minute average Note: If the required dose projections cannot be completed within the 60 minute period, then the declaration must be made based on the valid sustained monitor reading.

MODE All BASIS The term "Unplanned", as used in this context, includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

Unplanned releases in excess of 0.114 mRem/hr TPARD or 0.342 mRem/hr CDE that continue for > 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.

It is not intended that the release be averaged over 60 minutes, but exceed 0.114 mRem/hr TPARD or 0.342 mRem/hr CDE limits for 60 minutes. This EAL includes a 60 minute average for the dose projection with the release point radiation monitor above two times the HiHi alarm set point value for the entire 60 minutes. Also, it is intended that the event be declared as soon as it is determined that the release will exceed 0.114 mRem/hr TPARD or 0.342 mRem/hr CDE for greater than 60 minutes.

An indication or report is considered to be valid when it is verified by:

1. An instrument channel check

2. Indications on related or redundant instruments

3. By direct observation by plant personnel

ERP-101 BASES, Rev 3 Page 67 of 133 Monitor indications are calculated based on the methodology of the site Offsite Dose Calculation Manual (ODCM). The HiHi alarm setpoints are set conservatively to indicate when a potential release may approach Technical Specification (ODCM) limits assuming multiple release points. Use of this conservative setpoint in establishing a monitor reading will not cause an inappropriate event classification since this EAL requires the magnitude of the monitor reading to be two times the setpoint, sustained for >60 minutes, and assessment by a dose projection indicating an offsite dose rate in excess of two times Technical Specification (ODCM) limits. In the unlikely event that a dose projection cannot be completed within the 60 minute period, the event will be declared based on the sustained monitor reading.

Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.

The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly allowable Technical Specification limit (500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 2 times Technical Specifications [ODCM].

TPARD = 2x(Tech Spec Limit)/(hours per year)

= 2(500 mRem/yr.)/(8760 hr/yr.)

= 0.114 mRem/hr The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 2 times Technical Specifications [ODCM].

CDE = 2x(Tech Spec Limit)/(hours per year)

= 2(1500 mRem/yr.)/(8760 hr/yr.)

= 0.342 mRem/hr This event will be escalated to an Alert when effluents increase.

DEVIATION None REFERENCES NUMARC NESP-007, AU1.1 Offsite Dose Calculation Manual NUMARC Questions and Answers, June 1993, "Abnormal Rad Levels/Radiological Effluents

1. 9"

ERP-101 BASES, Rev 3 Page 68 of 133 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT - 5.1.1.b IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times Radiological Technical Specifications for 60 Minutes or Longer EAL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO TIMES Tech Specs (Liquid Release ODCM. 3.8.B.1 and Gaseous Release ODCM 3.8.C.1 .b) for > 60 minutes MODE All BASIS Releases in excess of two times technical specifications that continue for > 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety.

The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.

It is not intended that the release be averaged over 60 minutes, but exceed two times technical specifications limits for 60 minutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two times technical specifications for greater than 60 minutes.

An indication or report is considered to be valid when it is verified by:

1. An instrument channel check

2. Indications on related or redundant instruments

3. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).

This event will be escalated to an Alert when effluents increase.

DEVIATION None REFERENCES NUMARC NESP-007 AU1.2 Offsite Dose Calculation Manual T-104, Radioactivity Release Control

ERP-101 BASES, Rev 3 Page 69 of 133 5.0 Radioactivity Release 5.1 Effluent Release and Dose ALERT - 5.1.2.a IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL A valid reading on one or more of the following radiation monitors that exceeds TWO HUNDRED TIMES the HiHi alarm setpoint value for > 15 minutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mRem/hr TPARD OR 34.2 mRem/hr child thyroid CDE based on a 15 minute average Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

MODE All BASIS Releases in excess of 11.4 mRem/hr TPARD or 34.2 mRem/hr CDE that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event] and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.

This EAL includes a 15 minute average for the dose projection with the release point radiation monitor above two hundred times the HiHi alarm set point value for the entire 15 minutes.

Also, it is intended that the event be declared as soon as it is determined that the release will exceed 11.4 mRem/hr TPARD or 34.2 mRem/hr CDE for greater than 15 minutes.

An indication or report is considered to be valid when it is verified by:

1. An instrument channel check

2. Indications on related or redundant instruments

3. By direct observation by plant personnel Monitor indications are calculated based on the methodology of the site Offsite Dose Calculation Manual (ODCM). The HiHi alarm setpoints are set conservatively to indicate when a potential release may approach Technical Specification (ODCM) limits assuming multiple release points. Use of this conservative setpoint in establishing a monitor reading will not cause an inappropriate event classification since this EAL requires the magnitude of the monitor reading to be two hundred times the setpoint, sustained for >15 minutes, and assessment by a dose projection indicating an offsite dose rate in excess of two hundred times Technical Specification (ODCM) limits. In the unlikely event that a dose projection cannot be

ERP-101 BASES, Rev 3 Page 70 of 133 completed within the 15 minute period, the event will be declared based on the sustained monitor reading.

Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.

The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly allowable Technical Specification limit (500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 200 times Technical Specifications [ODCM].

TPARD = 200x(Tech Spec Limit)/(hours per year)

= 200(500 mRem/yr.)/(8760 hr/yr.)

= 11.4 mRem/hr The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 200 times Technical Specifications [ODCM].

CDE = 200x(Tech Spec Limit)/(hours per year)

= 200(1500 mRem/yr.)/(8760 hr/yr.)

= 34.2 mRem/hr This event will be escalated to a Site Area Emergency when actual or projected doses are determined to exceed 10CFR20 annual average population exposure limits.

DEVIATION None REFERENCES NUMARC NESP-007 AA1.1 Offsite Dose Calculation Manual NUMARC Questions and Answers, June 1993, "Abnormal Rad Levels/Radiological Effluents

1. 9"

ERP-101 BASES, Rev 3 Page 71 of 133 5.0 Radioactivity Release 5.1 Effluent Release and Dose ALERT - 5.1.2.b IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO HUNDRED TIMES Tech Specs (Liquid Release ODCM.

3.8.B.1 and Gaseous Release ODCM 3.8.C.1.b) for > 15 minutes MODE All BASIS Releases in excess of two hundred times technical specifications that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event] and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.

It is not intended that the release be averaged over 15 minutes, but exceed two hundred times technical specifications limits for 15 minutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two hundred times technical specifications for greater than 15 minutes.

An indication or report is considered to be valid when it is verified by:

1. An instrument channel check

2. Indications on related or redundant instruments

3. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).

This event will be escalated to higher classifications based on plant conditions.

DEVIATION None REFERENCES NUMARC NESP-007 AA1.2 Offsite Dose Calculation Manual T-104, Radioactivity Release Control

ERP-101 BASES, Rev 3 Page 72 of 133 5.0 Radioactivity Release 5.1 Effluent Release and Dose SITE AREA EMERGENCY - 5.1.3 IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release EAL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:

Main Stack 5.84 pCi/cc Vent Stack 2.08E-3 gCi/cc Torus Vent 203 cpm Note: If the dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

OR Projected offsite dose using computer dose model exceeds 100 mRem TPARD OR 500 mRem child thyroid CDE OR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mRem/hr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 500 mRem for one hour of inhalation MODE All BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

A monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or,

3. Direct observation by plant personnel.

Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.

ERP-101 BASES, Rev 3 Page 73 of 133 An actual or projected dose of 100 mrem Total Protective Action Recommendation Dose (TPARD) is based on the 10 CFR 20 annual average population exposure limit. This value also provides a desirable gradient (one order of magnitude) between the Site Area Emergency and General Emergency classifications. The 500 mrem integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TPARD and Child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used, since it gives the most accurate dose projection.

Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorology and a one hour release duration. The inputs are as follows:

Main Stack Vent Stack Torus Vent Stability Class E E E Wind Speed 11.4 mph 6.3 mph 6.3 mph Wind Direction 450 220 220 Accident LOCA LOCA LOCA Release Rate 5.84 iiCi/cc 2.08E-3 pCi/cc 203 cpm Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).

This event will be escalated to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines per EAL Section 5.1.4.

DEVIATION None REFERENCES NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400

ERP-101 BASES, Rev 3 Page 74 of 133 5.0 Radioactivity Release 5.1 Effluent Release and Dose GENERAL EMERGENCY - 5.1.4 IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology EAL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:

Main Stack 58.4 pCi/cc Vent Stack 2.08E-2 RCi/cc Torus Vent 2000 cpm Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

OR Projected offsite dose using computer dose model exceeds 1000 mRem TPARD OR 5000 mRem child thyroid CDE OR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mRem/hr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mRem for one hour of inhalation MODE All BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

A monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or,

3. Direct observation by plant personnel.

Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.

ERP-101 BASES, Rev 3 Page 75 of 133 The 1000 mR TPARD and the 5000 mR child thyroid integrated dose are based on the EPA protective action guidance. This is consistent with the emergency class description for a General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency. Actual meteorology is specifically identified in the initiating condition since it gives the most accurate dose assessment.

Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorology and a one hour release duration. The inputs are as follows:

Main Stack Vent Stack Torus Vent Stability Class E E E Wind Speed 11.4 mph 6.3 mph 6.3 Wind Direction 450 220 220 Accident LOCA LOCA LOCA Release Rate 58.4 pCi/cc 2.08E-2 gCi/cc 2.026E+3 cpm Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection (BRP).

DEVIATION None REFERENCES NUMARC NESP-007, AG1.1, AGI.3 and AG1.4 EPA-400

ERP-101 BASES, Rev 3 Page 76 of 133 5.0 Radioactivity Release 5.2 In-Plant Radiation UNUSUAL EVENT - 5.2.1 IC Unexpected Rise in Plant Radiation or Airborne Concentration EAL Valid Direct Area Radiation Monitor readings rise by a factor of 1000 over normal* levels Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

MODE All BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

An area monitor reading is considered to be valid when it is verified by:

1. an instrument channel check indicating the monitor has not failed;

2. indications on related or redundant instrumentation; or

3. direct observation by plant personnel This EAL addresses unplanned increases in in-plant radiation levels that represent a degradation in the control of radioactive material, and represents a potential degradation in the level of safety of the plant.

This event will be escalated to an Alert when radiation levels increase in areas required for the safe shutdown of the plant resulting in impeded access.

DEVIATION None REFERENCES NUMARC NESP-007, AU2.4 T- 103, Secondary Containment Control

ERP-101 BASES, Rev 3 Page 77 of 133 5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.a IC Release of Radioactive Material or Rises in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Valid radiation level readings > 5000 mR/hr in areas requiring infrequent access to maintain plant safety functions as identified in procedure SE-1 or SE-10 AND Access is required for safe plant operation, but is impeded, due to radiation dose rates MODE All BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

An area monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or,

3. Direct observation by plant personnel.

The single value of 5000 mR/hr was selected because it is based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e., 10 CFR 20), and in doing so, will impede necessary access. Stay times for levels up to that value are, generally several minutes, enough time to enter an area and manually operate the equipment.

This EAL addresses increased radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually, in order to maintain safe operation or perform a safe shutdown. These areas are identified in procedures SE-1 and SE-10. Use of these procedures will indicate the need to access the areas. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the increase in radiation levels is not a concern of this IC. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other IC may be involved. For example, a dose rate of 15 mR/hr in the control room or hi radiation monitor readings may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a SAE or GE may be indicated by the fission product barrier table.

This EAL could result in declaration of an Alert at one unit due to a radioactivity release or radiation shine resulting from a major accident at the other unit.

ERP-101 BASES, Rev 3 Page 78 of 133 This EAL is not meant to apply to increases in drywell radiation monitors, as these are events which are addressed in the fission product barrier table. Nor is it intended to apply to anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, etc.)

This event will be escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses.

DEVIATION None REFERENCES NUMARC NESP-007, AA3.2 T-103, Secondary Containment Control SE-1, Plant Shutdown from the Remote Shutdown Panel SE-10, Plant Shutdown from the Alternative Shutdown Panels

ERP-101 BASES, Rev 3 Page 79 of 133 5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.b IC Release of Radioactive Material or Rises in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Valid Control Room OR Central Alarm Station radiation reading > 15 mRlhr MODE All BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

An area monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or,

3. Direct observation by plant personnel.

The EAL address radiation levels which would impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey.

Plant normal and emergency procedures may be implemented without requiring any areas except the Control Room and Central Alarm Station to be continuously occupied. The Radwaste Control Room is not required to be continuously occupied in order to maintain plant safety functions since inputs to radwaste will be isolated with a secondary containment isolation and releases can only be performed manually.

The value of 15 mR/hr is derived from the GDC 19 value of 5 REM in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.

This event will be escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses.

ERP-101 BASES, Rev 3 Page 80 of 133 DEVIATION None REFERENCES NUMARC NESP-007 AA3.1

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ERP-101 BASES, Rev 3 Page 82 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power UNUSUAL EVENT - 6.1.1.a IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EAL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND At least Two Diesel Generators are supplying power to their respective 4 KV emergency busses MODE All BASIS This EAL addresses the loss of offsite AC power supplying the station. Offsite power is fed through 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer.

Loss of offsite power will cause a reactor scram and a containment isolation. All four (4) emergency Diesel Generators will be available to carry the essential loads for each unit (the four Diesel Generators are shared between each unit). Balance of Plant systems that would assist in plant operations (i.e., condensate pumps, etc.) may be unavailable due the loss of power.

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Implementation of this EAL is based on the number of powered 4 KV buses per unit.

Escalation of this event to an Alert would be based on having a loss of all offsite AC power coincident with onsite AC power being reduced to a single power source in Modes 1, 2, and 3 or having a loss of all offsite and onsite AC power in Modes 4 or 5.

DEVIATION None REFERENCES NUMARC NESP-007, SUW SE-1 1, Station Blackout

ERP-101 BASES, Rev 3 Page 83 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power UNUSUAL EVENT - 6.1.1.b IC Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes EAL The following conditions exist:

Unplanned Loss of ALL safety related DC Power indicated by < 107.5 VDC bus voltage indications for DC Panels 2(3)0D21, 22, 23, 24 AND Failure to restore power to at least one required DC bus within 15 minutes from the time of the loss MODE 4,5 BASIS The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. The safety related 125 volt DC Distribution Panels are as follows:

Unit 2 Unit 3 20D21 30D21 20D22 30D22 20D23 30D23 20D24 30D24 107.45 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. The value of 107.5 VDC will be used for human factors concerns.

This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is near the minimum voltage selected when battery sizing is performed.

Unplanned is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely, plants will perform maintenance on a Train related basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will occur.

ERP-101 BASES, Rev 3 Page 84 of 133 DEVIATION None REFERENCES NUMARC NESP-007, SU7 SE-13, Loss of a 125/250 VDC Safety Related Bus

ERP-101 BASES, Rev 3 Page 85 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.a IC AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout EAL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND Only One 4 KV emergency bus powered from a Single Onsite Power Source due to the Loss of: Three of Four Division Diesel Generators, D/G Output Breakers, or 4 KV Emergency Busses as indicated by bus voltage MODE 1, 2, 3 BASIS This EAL is intended to provide an escalation from "Loss of offsite Power for greater than 15 minutes." This condition is a degradation of the offsite and onsite power systems such that any additional failure would result in a station blackout. Fifteen (15) minutes has been selected to exclude transient or momentary power losses. However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.

Depending on the 4 KV AC bus that remains energized there is a disparity in the systems that may be available. The ability to remove heat from the containment via Torus cooling may be lost due to the need to operate the remaining available RHR pump in other than Torus cooling (e.g., LPCI). As such there is a decrease in the systems available to remove heat transferred to the containment and there is an ongoing release of energy from the reactor to the containment (via SRVs, HPCI and/or RCIC operation). The ability to cool the nuclear fuel, remove decay heat, and control containment parameters is severely limited. Should equipment be unavailable prior to the loss of power, functions necessary to maintain the plant in a cold shutdown condition may be threatened.

Implementation of this EAL is based on the number of powered 4 KV buses per unit.

Escalation of this event would be based on the loss of the remaining Emergency Diesel Generator.

DEVIATION None

ERP-101 BASES, Rev 3 Page 86 of 133 REFERENCES NUMARC NESP-007, SA5 SE-1 1, Station Blackout

ERP-101 BASES, Rev 3 Page 87 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.b IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode EAL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power MODE 4,5, D BASIS Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown, refueling, or defueled mode, the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency, if appropriate, is be Effluent Release/In-Plant Radiation, or Emergency Director Judgment.

Fifteen (15) minutes has been selected to exclude transient or momentary power losses.

However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.

Implementation of this EAL is based on the number of powered 4 KV buses per unit.

DEVIATION None REFERENCES NUMARC NESP-007, SA1 SE-11, Station Blackout

ERP-101 BASES, Rev 3 Page 88 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.a IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EAL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC MODE 1, 2,3 BASIS Control Room annunciators would indicate that all offsite and onsite AC power feeds have been lost. Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, High Pressure Service Water, and Emergency Service Water. Although instrumentation (supplied through instrument inverters) and DC power loads would be available, their operability would be limited to the amount of stored energy contained in their respective batteries. Instrumentation, communication equipment, and in-plant lighting and ventilation will be significantly hampered by the loss of all AC power.

Fifteen (15) minutes has been selected to exclude transient or momentary power losses.

However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.

Implementation of this EAL is based on the number of powered 4 KV buses per unit.

Escalation of this event would be based on the time that the Emergency Diesel Generator are unavailable.

DEVIATION None REFERENCES NUMARC NESP-007, SS1 SE-11, Station Blackout

ERP-101 BASES, Rev 3 Page 89 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.b IC Loss of All Vital DC Power EAL Loss of ALL Safety Related DC Power indicated by < 107.5 VDC on DC Panels 2(3)0D21, 22, 23, 24 for > 15 minutes MODE 1, 2, 3 BASIS:

A loss of all DC power compromises the ability to monitor and control plant functions. 125 Volt DC system provides control power to engineered safety features valve actuation, diesel generator auxiliaries, plant alarm and indication circuits as well as the control power for the associated load group. If 125 Volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions such as RPS Logic, Alternate Rod Insertion, Emergency Service Water Indication, 4KV Breaker Controls, HPCI, RCIC and RHR pump controls required to maintain safe plant conditions may not operate and core uncovery with subsequent reactor coolant system and primary containment failure might occur. The 125 volt DC Main Distribution Panel Busses are as follows:

Unit. 2 Unit 3 20D21 30D21 20D22 30D22 20D23 30D23 20D24 30D24 Loss of all DC Power causes the loss of the following equipment:

• Alternate Rod Insertion 0 ADS

• HPCI 0 RCIC

"* Normal EDG Control 0 Normal Recirculation Pump Trip

"* Containment Instrument Gas Compressors

"* Other 4KV Circuit Breakers (e.g., RHR, CS, CRD)

Loss of ADS creates a loss of low pressure ECCS due to the inability to depressurize the reactor. In addition, loss of these buses will eventually lead to MSIV closure and reactor trip due to the loss of the Containment Instrument Gas Compressor as a result of suction valve closure. Subsequent to MSIV closure, much of the equipment noted above will be required for plant stabilization and shutdown.

ERP-101 BASES, Rev 3 Page 90 of 133 A sustained loss of DC power will threaten the ability to remove heat from the reactor core, resulting in eventual fuel clad damage. The loss of DC power will also result in the loss of the ability to remove heat from the containment. SRVs will remain operable in the relief mode and the heat addition to the containment could result in a loss of the primary containment as a fission product release barrier.

107.45 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This EAL uses 107.5 VDC for human factors concerns. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is near the minimum voltage selected when battery sizing is performed.

DEVIATION None REFERENCES NUMARC NESP-007, SS3 T-101, RPV Control T-102, Primary Containment Control SE-1 1, Station Blackout

ERP-101 BASES, Rev 3 Page 91 of 133 6.0 Loss of Power 6.1 Loss of AC or DC Power GENERAL EMERGENCY - 6.1.4 IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EAL Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure of ALL Emergency Diesel Generators to supply power to 4 KV emergency busses AND At least one of the following conditions exist:

• Restoration of at least One 4 KV emergency bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is NOT likely OR

• Reactor Water Level cannot be maintained > -172" OR

• Torus temperature is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-102, T/T-1)

MODE 1, 2,3 BASIS When evaluating this EAL for Torus level outside of the Heat Capacity Temperature Limit Curve, High or Low, it is appropriate to consider operation to be on the "UNSAFE" side.

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. The two hours to restore AC power is based on the site blackout coping analysis as described below. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

10 CFR 50.2 defines Station Blackout (SBO) as complete loss of AC power to essential and non-essential buses. SBO does not include loss of AC Power to busses fed by station batteries through inverters, nor does it assume a concurrent single failure or design basis accident. Successful SBO coping maintains the following key parameters within given acceptable limits:

1. Reactor water level > -172" (TAF)

2. Torus level low enough to prevent HPCI and/or RCIC steam exhaust line flooding

3. Reactor pressure >150 psig to maintain HPCI and RCIC operable

ERP-101 BASES, Rev 3 Page 92 of 133

4. Containment pressure < 60 psig, design limit

5. Torus temperature < 200 degrees F, HPCI/RCIC lube oil temperature concern when suction aligned to Torus

6. Drywell temperature

<200 degrees F indefinitely

<250 degrees F 99 days

<320 degrees F 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />

<340 degrees F 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Successful extended SBO coping depends on ability to keep HPCI/RCIC available for injection, and ability to maintain RPV depressurized for low pressure injection should HPCI and RCIC become unavailable. Control power for HPCI, RCIC and SRVs is provided by 125V DC.

The parameters listed above can be maintained as long as the batteries are intact. Two hours is the earliest the batteries would fail, and thus is the basis for the time limit in this EAL.

The significance of a station blackout relative to the loss of fission product release barriers is that all three barriers will eventually be lost due to the inability to remove heat from the fuel and the containment. Although the RCS will be intact the longest, eventually SRVs will operate in the relief mode due to RPV over-pressurization and if the containment has already failed then there is a direct bypass of the RCS boundary.

Implementation of this EAL is based on the number of powered 4 KV buses per unit.

DEVIATION None REFERENCES NUMARC NESP-007, SG1 SE-1 1, Station Blackout T-101, RPV Control T-102, Primary Containment Control T-104, Radioactivity Release Control

ERP-101 BASES, Rev 3 Page 93 of 133 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation UNUSUAL EVENT - 7.1.1 IC Inability to Reach Required Shutdown Within Technical Specification Limits EAL Inability to reach required shutdown mode within Tech. Spec. LCO required action completion time.

MODE 1, 2, 3 BASIS Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a one hour report under 10 CFR 50.72 (b) Non-emergency events.

The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Notification of an Unusual Event is required when it is determined that the plant cannot be brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, SU2 Technical Specifications

ERP-101 BASES, Rev 3 Page 94 of 133 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation ALERT - 7.1.2 IC Control Room Evacuation Has Been Initiated EAL Entry into SE-1 or SE-10 procedure for Control Room evacuation MODE All BASIS Control Room evacuation requires establishment of plant control from outside the control room (e.g., local control and remote shutdown panel) and support from the Technical Support Center and/or other emergency facilities as necessary. Control Room evacuation represents a serious plant situation since the level of control is not as complete as it would be without evacuation. The establishment of system control outside of the Control Room will bypass many protective trips and interlocks. In addition, much of the instrumentation and assessment tools available in the Control Room will not be available.

This event will be escalated to a Site Area Emergency if control cannot be established within fifteen minutes.

DEVIATION None REFERENCES NUMARC NESP-007, HA5 SE-10, Alternate Shutdown SE-1 Plant Shutdown from the Remote Shutdown Panel

ERP-101 BASES, Rev 3 Page 95 of 133 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation SITE AREA EMERGENCY - 7.1.3 IC Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EAL The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-1 or SE-10 within 15 minutes MODE All BASIS Transfer of safety system control has not been performed in an expeditious manner but it is unknown if any damage has occurred to the fission product barriers. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the control room, arrive at the remote shutdown area, and reestablish plant control to preclude core uncovery and/or core damage. During this transitional period the function of monitoring and/or controlling parameters necessary for plant safety may not be occurring and as a result there may be a threat to plant safety.

This event will be escalated based upon system malfunctions or damage consequences.

DEVIATION None REFERENCES NUMARC NESP-007, HS2 SE-IO, Alternate Shutdown SE-1, Plant Shutdown from the Remote Shutdown Panel

ERP-101 BASES, Rev 3 Page 96 of 133 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability ALERT - 7.2.2 IC Inability to Maintain Plant in Cold Shutdown EAL The following conditions exist:

Unplanned Loss of ALL Tech Spec required systems available to provide Decay Heat Removal functions AND Uncontrolled Temperature rise that either:

• Exceeds 212 OF (Excluding a <15 minute rise >2120 F with a heat removal function restored)

OR

• Results in temperature rise approaching 212 OF (with NO heat removal function restored)

MODE 4, 5 BASIS This EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. A loss of Technical Specifications components is paired with exceeding temperature limits to acknowledge additional plant capabilities to maintain plant cooling. Escalation to Site Area Emergency or General Emergency would be via Effluent Release/In-Plant Radiation or Emergency Director Judgment ICs.

The statement "Unplanned Loss of ALL Tech Spec required systems available to provide Decay Heat Removal functions" is intended to represent a complete loss of functions available, or an inadequate ability, to provide core cooling during the Cold Shutdown and Refueling Modes, including alternate decay heat removal methods. This EAL allows for actions taken in ON-125, "Loss of Shutdown Cooling - Procedure," to reestablish RHR in the Shutdown Cooling Mode or provide for alternate methods of decay heat removal, with the intent of maintaining RCS temperature below 2120 F.

For loss of an in-service Decay Heat Removal system with other decay heat removal methods available, actions taken to provide for restoration of a decay heat removal function may require time to implement. If the event results in RCS temperature "momentarily" (for less than 15 minutes) rising above 212OF with heat removal capability restored, Emergency Director/Shift Management judgment will be required to determine whether heat removal systems are adequate to prevent boiling in the core and restoration of RCS temperature control.

Momentary (not to exceed 15 minutes) unplanned excursions above 2120 F, when alternate decay heat removal capabilities exist, should not be classified under this EAL.

ERP-101 BASES, Rev 3 Page 97 of 133 "Uncontrolled" means that system temperature rise is not the result of planned actions by the plant staff.

The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.

This EAL is concerned with the ability to keep the reactor core temperature less than 212 °F.

The criteria of uncontrolled Reactor Coolant temperature rise > 212 OF is met as soon as it becomes known that sufficient cooling cannot be restored in time to maintain the temperature

< 212 OF, regardless of the current temperature. The inability to establish alternate methods of decay heat removal indicates that either alternate methods are unavailable to cool the core in the RPV or when the steam is transferred to the Torus, Torus cooling is unavailable. Loss of Torus cooling will result in a continuing, uncontrolled increase in reactor coolant temperature.

Escalation to the Site Area Emergency is by EAL IC, "Loss of Water Level in the Reactor Vessel that has or will uncover Fuel in the Reactor Vessel," or by Effluent Release/In-Plant Radiation ICs.

DEVIATION None REFERENCES NUMARC NESP-007, SA3 ON-125, Loss of Shutdown Cooling - Procedure Technical Specifications

ERP-101 BASES, Rev 3 Page 98 of 133 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability SITE AREA EMERGENCY - 7.2.3 IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EAL Loss of TORUS heat sink capabilities as evidenced by T-1 02 T/T legs directing a T-1 12 Emergency Blowdown MODE 1, 2,3 BASIS:

This EAL is concerned with Torus temperature. It is not appropriate to make a Site Area Emergency classification for the condition where the T -102 Torus Level leg alone directs a T-1 12 Emergency Blowdown since the Emergency Blowdown is performed PRIOR to those Torus levels which may cause a loss of containment capability due to uncovering downcomers or excessive SRV tailpipe stresses.

This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other EALs. The loss of heat removal function is indicated by T-102 T/T legs requiring an Emergency Blowdown which is directed when the Heat Capacity Temperature Limit (HCTL) curve is exceeded.

Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency would be via Effluent Release/In-Plant Radiation, Emergency Director Judgment, or Fission Product Barrier Degradation ICs.

DEVIATION None REFERENCES NEI 97-03, SSA T-102, Primary Containment Control, SP/L-8

ERP-101 BASES, Rev 3 Page 99 of 133 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.a IC Unplanned Loss of Most or All Safety System Annunciation or Indication in The Control Room for Greater Than 15 Minutes EAL Unplanned loss of most or all safety system annunciators (Table 7-1) OR indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit(s).

MODE 1, 2,3 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Supervisor this loss of annunciators requires increased surveillance to safely operate the plant. It is not intended that a detailed count of instrumentation be performed, but that only a rough approximation be used to determine the severity of the loss. The Plant Monitoring System (PMS) is available to provide compensatory indication. Fifteen minutes is used as a threshold to exclude transient or momentary power losses. Unplanned loss of annunciators excludes scheduled maintenance and testing activities. Control Room panels with annunciators and direction for response are included in ON-123, Loss of Control Room Annunciators.

Table 7-1 indicates those system annunciator panels considered to be safety related:

Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 indicates those indications important for monitoring:

Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in EAL section, Technical Specifications.

This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.

This event will be escalated to an Alert if a transient is in progress or if compensatory indications become unavailable.

ERP-101 BASES, Rev 3 Page 100 of 133 DEVIATION None REFERENCES NUMARC NESP-007, SU3 ON-123, Loss of Control Room Annunciators AIT A0004447, EP Self Assessment on Salem Loss of Annunciators

ERP-101 BASES, Rev 3 Page 101 of 133 7.0 Internal Events 7.3 Loss of Assessment I Communication Capability UNUSUAL EVENT - 7.3.1.b IC Unplanned Loss of All Onsite or Offsite Communications Capabilities EAL MODE All BASIS This EAL recognizes a loss of communication ability that significantly degrades the plant operations staff's ability to perform tasks necessary for plant operations or the ability to communicate with offsite authorities. This EAL is separated into two groups of communications, Onsite and Offsite. A complete loss of either group is so severe, that the Unusual Event declaration is warranted. Table 7-2 is identified as follows:

Table 7-3 Communications Onsite Offsite Site Phones (GTE System) x x OMNI System x x Plant Public Address x Station Radio x NRC (FTS-2000) x PA State Police Radio x Load Dispatcher Radio x PECO Dial Network x There is no escalation to an Alert for loss of communications, although there is escalation to higher classifications if other communications for plant assessment is lost.

DEVIATION None REFERENCES NUMARC NESP-007, SU6 Nuclear Emergency Plan

ERP-101 BASES, Rev 3 Page 102 of 133 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability ALERT - 7.3.2 IC Unplanned Loss of Most or All Safety System Annunciation or Indication In Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non Alarming Indicators are Unavailable EAL Unplanned loss of most or all safety system annunciators (Table 7-1) OR indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit(s)

AND EITHER A significant plant transient is in progress (Table 7-4) OR the plant monitoring system (PMS) is unavailable.

MODE 1, 2,3 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Supervisor this loss of annunciators requires increased surveillance to safely operate the plant. This EAL represents an increase in severity above 7.3.1.a in that the Plant Monitoring System (PMS) can not provide compensatory indication, or that a significant transient is in progress.

Table 7-1 indicates those system annunciator panels considered to be safety related:

Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 indicates those indications important for monitoring:

Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions

ERP-101 BASES, Rev 3 Page 103 of 133 Table 7-4, significant plant transients include response to automatic or manually initiated actions including:

Table 7-4 Plant Transients SCRAM Recirc runbacks > 25% thermal power Sustained power oscillations 25% peak to peak Stuck open relief valves ECCS injection Fifteen minutes is used as a threshold to exclude transient or momentary power loses.

Control Room panels with annunciators and direction for restoration is included in ON-123, Loss of Control Room Annunciators.

Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in EAL section, Technical Specifications.

This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.

This event will be escalated to a Site Area Emergency if a transient is in progress, the Plant Monitoring System is unavailable and a loss of annunciators occurs.

DEVIATION None REFERENCES NUMARC NESP-007, SA4 ON-123, Loss of Control Room Annunciators T-101, Bases BWROG EPG/SAG (RC/Q-6)

ERP-101 BASES, Rev 3 Page 104 of 133 7.0 Internal Events 7.3 Loss of Assessment I Communication Capability SITE AREA EMERGENCY - 7.3.3 IC Inability to Monitor a Significant Transient in Progress EAL Loss of safety system annunciators (Table 7-1)

AND indicators (Table 7-2)

AND PMS AND a significant plant transient is in progress. (Table 7-4)

MODE 1, 2,3 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Supervisor this loss of annunciators requires increased surveillance to safely operate the plant. This EAL represents an increase in severity above 7.3.2 in that the Plant Monitoring System can not provide compensatory indication, and that a significant transient is in progress.

Table 7-1 indicates those system annunciator panels considered to be safety related:

Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 indicates those indications important for monitoring:

Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions Table 7-4 significant plant transients include response to automatic or manually initiated actions including:

Table 7-4 Plant Transients SCRAM Recirc runbacks >25% thermal power change Sustained power oscillations 25% peak to peak Stuck open relief valves ECCS injection

ERP-101 BASES, Rev 3 Page 105 of 133 Planned maintenance or testing activities are included in this EAL due to the significance of this event. Control Room panels with annunciators and the restoration is included in ON-123, Loss of Control Room Annunciators.

DEVIATION None REFERENCES NUMARC NESP-007, SS6 ON-123, Loss of Control Room Annunciators T-101, Bases BWROG EPG/SAG (RC/Q-6)

ERP-101 BASES, Rev 3 Page 106 of 133 8.0 External Events 8.1 Security Events UNUSUAL EVENT - 8.1.1 IC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EAL A credible threat to the station reported by the NRC.

OR An Actual Threat that meets ALL of the following criteria:

"* A credible threat reported by any other outside agency or determined per SY-AA 101-132; AND

", Is specifically directed towards the station; AND

"* Is imminent (within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)

OR Attempted intrusion and attack to the Protected Areas OR Attempted sabotage discovered within the Protected Areas OR Hostage/Extortion situation that threatens normal plant operations MODE All BASIS A security threat that is identified as being directed towards the station and represents a potential degradation in the level of safety of the plant. A security threat is satisfied if physical evidence supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The Shift Management will declare an Unusual Event subsequent to consulting with the on shift Security representative to determine the credibility of the security event.

Security threats which meet the threshold for declaration of an Unusual Event are:

1. A credible threat to the station reported by the NRC.

2. An Actual Threat that meets ALL of the following criteria:

"* A credible threat reported by any other outside agency or determined per SY-AA-101 132; AND

"* Is specifically directed towards the station; AND

"* Is imminent (within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)

3. Attempted intrusion and attack to the Protected Areas

4. Attempted sabotage discovered within the Protected Areas

5. Hostage/Extortion situation that threatens normal plant operations

ERP-101 BASES, Rev 3 Page 107 of 133 Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or 10 CFR 50.72 and will not cause an Unusual Event to be declared.

This event will be escalated to an Alert based upon a hostile intrusion or act within the Protected Areas.

DEVIATION A bomb device discovered within Plant Protected Areas and outside the Plant Vital Areas is an Alert declaration as determined per the site Safeguards Contingency Plan and therefore is not included as an Unusual Event in the EAL scheme.

REFERENCES NUMARC NESP-007, HU4.1 and HU4.2 Safeguards Contingency Plan Physical Security Plan

ERP-101 BASES, Rev 3 Page 108 of 133 8.0 External Events 8.1 Security Events ALERT - 8.1.2 IC Security Event in a Plant Protected Area EAL MODE All BASIS This class of security event represents an escalated threat to the level of safety of the plant.

This event is satisfied if physical evidence supporting the hostile intrusion or attack exists. The Shift Management will declare an Alert subsequent to consulting with the on shift Security representative to determine the validity of the entry conditions.

Security threats which meet the threshold for declaration of an Alert are:

1. Intrusion into plant protected areas by a hostile force

2. Confirmed bomb, sabotage or sabotage device discovered within the Protected Areas This event will be escalated to a Site Area Emergency based upon a hostile intrusion or act in plant Vital Areas.

DEVIATION None REFERENCES NUMARC NESP-007, HA4.1 and HA4.2 Safeguards Contingency Plan Physical Security Plan

ERP-101 BASES, Rev 3 Page 109 of 133 8.0 External Events 8.1 Security Events SITE AREA EMERGENCY - 8.1.3 IC Security Event in a Plant Vital Area EAL MODE All BASIS This class of security event represents an escalated threat to plant safety above that contained in an Alert in that a hostile intrusion or attack has progressed from the Protected Area to a Vital Area. The Vital Areas are within the Protected Area and are generally controlled by key card readers. These areas contain vital equipment which includes any equipment, system, device or material, the failure, destruction or release of could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect health and safety following such failure, destruction or release are also considered vital.

Security threats which meet the threshold for declaration of a Site Area Emergency are:

1. Intrusion into plant Vital area by a hostile force

2. Confirmed bomb, sabotage or sabotage device discovered in a Vital Area This event will be escalated to a General Emergency based upon the loss of physical control of the Control Room or Remote Shutdown Capability DEVIATION None REFERENCES NUMARC NESP-007, HSI.1 and HS1.2 Safeguards Contingency Plan Physical Security Plan

ERP-101 BASES, Rev 3 Page 110 of 133 8.0 External Events 8.1 Security Events GENERAL EMERGENCY - 8.1.4 IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown EAL Loss of physical control of the control room due to security event OR Loss of physical control of all remote shutdown capability due to security event MODE All BASIS This class of security event represents conditions under which a hostile force has taken physical control of areas required to reach and maintain cold shutdown. Loss of Remote Shutdown Capability would occur if the control function of the Remote Shutdown Panels was lost.

Security events which meet the threshold for declaration of a General Emergency are physical loss of the Control Room or the Remote and Alternate Shutdown Panels.

This situation leaves the plant in a very unstable condition with a high potential of multiple barrier failures.

DEVIATION None REFERENCES NUMARC NESP-007, HGI.1 and HG1.2 Safeguards Contingency Plan Physical Security Plan

ERP-101 BASES, Rev 3 Page 111 of 133 8.0 External Events 8.2 Fire I Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.a IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection EAL Fire within ON-1 14 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm MODE All BASIS The purpose of this IC is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This excludes such items as fires within administration buildings, waste-basket fires, and other small fires of no safety consequence.

This IC applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this IC is not to include buildings (e.g., warehouses) or areas that are not contiguous or immediately adjacent to plant vital areas. Verification of the alarm in this context means those actions taken in the control room to determine that the control room alarm is not spurious.

This EAL addresses fires in Plant Vital Structures that house safety systems. These fires may be precursors to damage to safety systems contained in these structures. There are no areas/buildings contiguous to Plant Vital Structures which could effect a safety system in one of the listed Plant Vital Structures except for those already on the list. Therefore, no additional areas/buildings are considered for this EAL. Verification that a fire exists is by operator actions to confirm that fire alarms received in the Control Room are not spurious or by any verbal notification by plant personnel. Fifteen minutes has been established to allow plant staff to respond and control small fires or to verify that no fire exists. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower This event will be escalated to an Alert if the fire damages redundant trains of plant safety systems required for the current operating condition.

DEVIATION None

ERP-101 BASES, Rev 3 Page 112 of 133 REFERENCES NUMARC NESP-007, HU2

ERP-101 BASES, Rev 3 Page 113 of 133 8.0 External Events 8.2 Fire I Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.b IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant EAL Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant OR Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event MODE All BASIS This EAL addresses toxic/flammable gas releases within the Protected Area in concentrations high enough to affect health of plant personnel or the safe operation of the plant. This includes releases that originate both onsite and offsite. A toxic/flammable gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. A gas release is considered to be impeding normal plant operations if concentrations are high enough to restrict normal operator movements. It also includes areas where access is only possible with respiratory equipment, as this equipment restricts normal visibility and mobility. It should not be construed to include confined spaces that must be ventilated prior to entry or situation involving the Fire Brigade who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with the Fire Brigade.

An offsite event (such as a tanker truck accident or train derailment releasing toxic gases) may place the Protected Area within the evacuation area. This evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials.

DEVIATION None REFERENCES NUMARC NESP-007, HU3.1 and HU3.2

ERP-101 BASES, Rev 3 Page 114 of 133 8.0 External Events 8.2 Fire / Explosion and Toxic I Flammable Gases UNUSUAL EVENT - 8.2.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment MODE All BASIS The protected area boundary is typically that part within the security isolation zone and is defined in the site security plan.

Only those explosions of sufficient force to damage permanent structures or equipment within the protected area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching) is sufficient for declaration. The Emergency Director also needs to consider any security aspects of the explosion, if applicable.

Any security aspects of this event should be considered under EAL Section 8.1, Security Events.

This event will be escalated to an Alert if the explosion damages one or more redundant trains of plant safety systems required for the current operating condition.

DEVIATION None REFERENCES NUMARC NESP-007, HUI.5

ERP-101 BASES, Rev 3 Page 115 of 133 8.0 External Events 8.2 Fire I Explosion and Toxic / Flammable Gases ALERT - 8.2.2.a IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL The following conditions exist:

Fire or explosion which potentially makes inoperable:

Two or More subsystems of a Safe Shutdown System (Table 8-2) OR Two or More Safe Shutdown Systems OR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Mode MODE All BASIS The primary concern of this EAL is the magnitude of the fire and the effects on Safe Shutdown Systems required for the present Operational Mode. A Safe Shutdown System is defined as any system required to maintain safe operation or to establish or maintain Cold Shutdown. A system being "inoperable" means that it is incapable of performing the design function. For example, the LPCI System is intended to maintain adequate core cooling by covering the core to at least 2/3 core height following a DBA LOCA. In order for the system to be unable to maintain its intended function, multiple loops would need to be disabled by the fire. In addition to indication of degraded system performance, potential inoperability may be determined by visual observation and other control room indications such as loss of indicating lights.

Table 8-2 Safe Shutdown Systems Diesel Generators 4KV Safeguard Buses ADS HPCI RCIC RHR (All Modes)

Core Spray HPSW ESW SBGTS ECW CAC/CAD PCIS Control Room Ventilation Safe Shutdown Analysis is consulted to determine systems required for the applicable mode.

Two examples of applying this methodology are as follows:

- Diesel Generators and 4 KV Safeguard Buses The fire disables multiple Diesel Generators or 4 KV Safeguard Buses so that the number of emergency power systems available would be decreased to below what would be required to mitigate an accident under the current operating conditions.

ERP-101 BASES, Rev 3 Page 116 of 133 For 100% power, this could be conservatively interpreted as at least two Diesel Generators or 4 KV Buses disabled.

- RHR - LPCI Mode The fire disables multiple loops of LPCI so that adequate core submergence could not be assured following a DBA LOCA. For 100% power, this could also be conservatively interpreted as at least two loops disabled.

The EAL includes the condition that the fire must make "TWO OR MORE" subsystems or "TWO OR MORE" systems inoperable. In those cases where it is believed that the fire may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL if the structure houses or otherwise supports Safety Systems required for the present Operational Mode.

Degraded system performance or observation of damage that could degrade system performance is used as the indicator that the safe shutdown system was actually affected or made inoperable. A report of damage should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of damage. The occurrence of the fire or explosion with reports of damage (e.g., deformation, scorching) is sufficient for declaration.

Fire is defined as combustion characterized by the generation of heat and smoke. Sources of smoke such as overheated electrical equipment and slipping drive belts, for example, do not constitute fires. Observation of a flame is preferred, but is NOT required if large quantities of smoke and heat are observed.

This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA2 PBAPS Safe Shutdown Analysis NUMARC Questions and Answers, June 1993, "Hazards Question #7"

ERP-101 BASES, Rev 3 Page 117 of 133 8.0 External Events 8.2 Fire / Explosion and Toxic I Flammable Gases ALERT - 8.2.2.b IC Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Report or detection of toxic gases within Plant Vital Structures (Table 8-1) in concentrations that will be life threatening to plant personnel OR Report or detection of flammable gases within Plant Vital Structures (Table 8-1) in concentrations affecting the safe operation of the plant MODE All BASIS This EAL recognizes that toxic/flammable gases have entered Plant Vital Structures and are affecting safe operation of the plant by impeding operator access to the safety systems that must be operated manually in these structures. The cause and/or magnitude of the gas concentrations is not a concern, but rather that access is required to an area and is impeded.

Plant Vital Structures that must be accessed are as follows:

Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not contiguous or immediately adjacent to plant Vital Areas. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA3.1 and HA3.2

ERP-101 BASES, Rev 3 Page 118 of 133 8.0 External Events 8.3 Man-Made Events UNUSUAL EVENT - 8.3.1.a IC Destructive Phenomena Affecting the Protected Area EAL Vehicle crash within protected area boundary that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.

MODE All BASIS This EAL is intended to address such items as plane, helicopter, or train crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be escalated to Alert.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.4

ERP-101 BASES, Rev 3 Page 119 of 133 8.0 External Events 8.3 Man-Made Events UNUSUAL EVENT - 8.3.1.b IC Destructive Phenomena Affecting the Protected Area EAL Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

MODE All BASIS This EAL is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (e.g., lubricating oils) and gases (e.g., hydrogen) to the plant environs. Actual fires and flammable gas build up are appropriately classified via other EALs. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure or by the radiological releases and would be classified by the radiological lCs or Fission Product Barrier lCs.

Turbine failure of sufficient magnitude to cause observable damage to the turbine casing or seals of the turbine generator increases the potential for leakage of combustible fluids and gases (Hydrogen cooling) to the Turbine Building. The damage should be readily observable and should not require equipment disassembly to locate.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.6

ERP-101 BASES, Rev 3 Page 120 of 133 8.0 External Events 8.3 Man-Made Events ALERT - 8.3.2 IC Destructive Phenomena Affecting the Plant Vital Area EAL Vehicle crash affecting Plant Vital Structures (Table 8-1)

OR Turbine failure generated missiles result in any visible structural damage to or penetration of any Plant Vital Structures (Table 8-1)

MODE All BASIS This EAL address crashes of vehicles or missile impacts that have caused damage to Plant Vital Structures, and thus damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification. The evidence of damage is sufficient for declaration. A vehicle crash includes aircraft and large motor vehicles, such as a crane. Missile impacts including flying objects from offsite, onsite rotating equipment or turbine failure causing casing penetration. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.5 and HA1.6

ERP-101 BASES, Rev 3 Page 121 of 133 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL Earthquake >.01 g as determined by procedure SO 67.7.A MODE All BASIS This EAL addresses a sensed earthquake. The magnitude of .Olg is the lowest detectable earthquake measured on PBAPS seismic instrumentation per SO 67.7.A. An earthquake of this magnitude may be sufficient to cause minor damage to plant structures or equipment within the Protected Area. Damage is considered to be minor, as it would not affect physical or structural integrity. This event is not expected to affect the capabilities of plant safety functions.

This event will be escalated to an Alert if the earthquake reaches an Operating Basis Earthquake.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.1 SE-5, Earthquake and Bases UFSAR, section 1.6

ERP-101 BASES, Rev 3 Page 122 of 133 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report by plant personnel of tornado striking within protected areas OR Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes MODE All BASIS A tornado touching down within the Protected Areas or wind speeds > 75 mph within the owner controlled Area are of sufficient velocity to have the potential to cause damage to Plant Vital Structures. The value of 75 mph was selected to maintain consistency with plant value and to coincide with the Beaufort Scale for Hurricane wind speed winds of 73-136 mph. These conditions are indicative of unstable weather conditions and represent a potential degradation in the level of safety of the plant. Verification of a tornado will be by direct observation and reporting by station personnel. Verification of wind speeds > 75 mph will be via meteorological data in the control room. For purposes of this EAL, sustained is > 15 minutes.

This event will be escalated to an Alert if the tornado or high wind speeds strike Plant Vital Structures. If it is determined that the tornado or high wind speeds have caused a loss of shutdown cooling, then escalation will be by EAL IC, Loss of Decay Heat Removal Capability.

DEVIATION None REFERENCES NUMARC NESP-007, HUI.2 and HU1.7

ERP-101 BASES, Rev 3 Page 123 of 133 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL Assessment by the control room that an event has occurred. (Natural and Destructive Phenomena Affecting the Protected Areas)

MODE All BASIS This EAL allows for the control room to determine that an event has occurred and take appropriate action based on personal assessment as opposed to verification (e.g., an earthquake is felt but does not register on any plant-specific instrumentation, etc.)

DEVIATION None REFERENCES NUMARC NESP-007, HU1.3

ERP-101 BASES, Rev 3 Page 124 of 133 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.d IC Natural and Destructive Phenomena Affecting the Protected Area EAL High River level > 112' OR Low River level <98.5' MODE All BASIS High River level of greater than 112 feet on instrument LI-2(3)278A,B,C or LI-2(3)278A,B,C is indication of the river being in flood. By procedure, the units will be SCRAMmed and be brought to cold shutdown.

Low River level of less than 98.5 feet is indication of loss of Conowingo Pond and loss of circulation water pumps. Procedures require the unit to be SCRAMmed and brought to cold shutdown.

This event will be escalated to an Alert classification based continuation of the river situation.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.7 SE-4, Flood SE-3, Loss of Conowingo Pond

ERP-101 BASES, Rev 3 Page 125 of 133 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.a IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Earthquake >.05 g (Operating Basis Earthquake OBE) as determined by procedure SO 67.7.A MODE All BASIS This EAL addresses an earthquake that exceeds the Operating Basis Earthquake level of .05g and is beyond design basis limits. An earthquake of this magnitude may be sufficient to cause damage to safety related systems and functions.

The Max Credible Earthquake for PBAPS is 0.12g per UFSAR section 1.6, therefore this EAL is conservative and warrants an Alert classification.

This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.1 SE-5, Earthquake and Bases UFSAR section 1.6

ERP-101 BASES, Rev 3 Page 126 of 133 8.0 External Events 8.4 Natural Events ALERT - 9.4.2.b IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 8-1)

MODE All BASIS This EAL is based on FSAR design basis. Wind loads of this magnitude can cause damage to safety functions.

This EAL addresses events where Plant Vital Structures have been struck with high winds, and thus damage may have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.2

ERP-101 BASES, Rev 3 Page 127 of 133 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.c IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Report of any visible structural damage to any Plant Vital Structure (Table 8-1)

MODE All BASIS The threshold value of this EAL should be determined relative to the damage that might occur from events described in EALs 8.4.2.a and 8.4.2.b.

This EAL specifies the Plant Vital Structures which contain systems and functions required for safe shutdown of the plant. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower Other site structures listed in the NUMARC document are not plant vital structures and are not required for safe shutdown. Those are: RWST, CST.

This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.3

ERP-101 BASES, Rev 3 Page 128 of 133 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.d IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL MODE All BASIS High River level > 116 feet is indication of the river being in flood. This level is capable of causing flooding that can affect Plant Vital Structures. No attempt should be made to determine the magnitude of flooding. This is a long lead time event but this level is ground elevation of the reactor building and intake pump structure so classification as an Alert Event is appropriate. The evidence of flooding is sufficient for declaration.

Low River level < 92.5 feet is indication of loss of Conowingo Pond and loss of circulation water pumps. Procedures require the unit to be SCRAMmed and brought to cold shutdown and utilization of the ECW pump and Emergency Cooling Tower.

This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.7 SE-4, Flood SE-3, Loss of Conowingo Pond

ERP-101 BASES, Rev 3 Page 129 of 133 This page intentionally left blank

ERP-101 BASES, Rev 3 Page 130 of 133 9.0 Other 9.1 General UNUSUAL EVENT - 9.1.1 IC Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Unusual Event EAL Other conditions exist which in the judgment of the Emergency Director indicate a potential degradation of the level of safety of the plant MODE All BASIS This EAL allows the Shift Management to declare an Unusual Event upon the determination that the level of safety of the plant has degraded. Where the degradation is associated with equipment or system malfunctions, the decision that it is degraded should be made upon functionality, not operability. A system, subsystem, train, component or device, though degraded in equipment condition or configuration, should be considered functional if it is capable of maintaining respective system parameters within acceptable design limits.

Releases of radioactive materials requiring offsite response or monitoring are not expected to occur at this level unless further degradation of safety systems occurs. However, if one does occur, it will be classified under "Radioactivity Releases."

DEVIATION None REFERENCES NUMARC NESP-007, HU5

ERP-101 BASES, Rev 3 Page 131 of 133 9.0 Other 9.1 General ALERT - 9.1.2 IC Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert EAL Other conditions exist which in the Judgment of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.

MODE All BASIS This EAL allows the Shift Management to declare an Alert upon the determination that the level of safety of the plant has substantially degraded but is not explicitly addressed by other EALs. This includes a determination by Shift Management that the TSC and OSC should be activated and command and control functions should be transferred for the event to be effectively mitigated. Transfer of command and control functions is used as an initiator since an event significant to warrant transfer is a substantial reduction in the level of safety of the plant. Other examples are:

Internal flooding affects the operability of plant safety systems required to establish or maintain cold shutdown.

Releases that are expected will be limited to a small fraction of the EPA Protective Action Guidelines and will be classified under "Radioactivity Releases."

DEVIATION None REFERENCES NUMARC NESP-007, HA6

ERP-101 BASES, Rev 3 Page 132 of 133 9.0 Other 9.1 General SITE AREA EMERGENCY - 9.1.3 IC Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency EAL Other conditions exist which in the Judgment of the Emergency Director indicate actual or likely major failures of plant functions needed for protection of the public MODE All BASIS This EAL allows the Shift Management to declare a Site Area Emergency upon the determination of an actual or likely major failure of plant functions needed for protection of the public, but is not explicitly addressed by other EALs.

Releases are not expected to result in exposure levels which exceed the EPA Protective Action Guidelines except within the site boundary and will be classified under "Radioactivity Releases."

DEVIATION None REFERENCES NUMARC NESP-007, HS3

ERP-101 BASES, Rev 3 Page 133 of 133 9.0 Other 9.1 General GENERAL EMERGENCY - 9.1.4 IC Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency EAL Other conditions exist which in the Judgment of the Emergency Director indicate: (1) actual or imminent substantial core degradation with potential for loss of containment, or (2) potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary MODE All BASIS This EAL allows the Shift Management to declare a General Emergency upon the determination of an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity, but is not explicitly addressed by other EALs.

Releases may exceed the EPA Protective Action Guidelines for more than the immediate site area and will be classified under "Radioactivity Releases."

DEVIATION None REFERENCES NUMARC NESP-007, HG2

ATTACHMENT 2 PEACH BOTTOM POWER STATION, UNITS 2 & 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 EMERGENCY RESPONSE PROCEDURES REPORT INDEX

2/04/2002 PAGE I PEACH BOTTOM ATOMIC POWER STATION PROCEDURE INDEX REPORT:

CURR DOC PROC REV EFFECTIVE RESP SYSTEM FAC TYPE TYPE PROCEDURE NUMBER NBR TITLE DATE GROUP NBR PB PROC ERP ERP-C-1000 0006 EMERGENCY OPERATIONS FACILITY (EOF) ACTIVATION/DECACTIVATION 06/25/01 PWE PB PROC ERP ERP-C-1000-1 0004 EOF ACTIVATION CHECKLIST 06/25/01 PWE PB PROC ERP ERP-C-1000-2 0003 EOF DEACTIVATION CHECKLIST 04/21/99 PWE PB PROC ERP ERP-C-1000-3 0000 EOF BUSINESS HOURS FIRST RESPONDER CHECKLIST 04/21/99 PWE PB PROC ERP ERP-C-1000-4 0000 EOF AFTER HOURS FIRST RESPONDER CHECKLIST 04/21/99 PWE PB PROC ERP ERP-C-1000-5 0000 MINIMUM STAFFING POSITIONS NECESSARY TO ACTIVATE THE EOF 06/25/01 PWE PB PROC ERP ERP-C-1100 0003 EOF STAFF AUGMENTATION- CANCELLED - REPLACED BY ERP-C-1250 09/14/94 PWE PB PROC ERP ERP-C-1200 0011 EMERGENCY REPSONSE MANAGER 06/25/01 PWE PB PROC ERP ERP-C-1200-1 0000 EMERGENCY RESPONSE MANAGER TURNOVER/BRIEFING FORM 09/14/94 PWE PB PROC ERP ERP-C-1200-2 EXH 0000 PROTECTIVE ACTION RECOMMENDATION WORKSHEET CANCELLED REPLACED BY 10/24/95 PWE ERP-C-1200 PB PROC ERP ERP-C-1200-3 0000 ERM PAR DELIVERY CHECKLIST 04/03/00 PWE PB PROC ERP ERP-C-1200-4 0000 MINIMUM STAFFING POSITIONS NECESSARY TO ACTIVATE THE EOF 03/30/01 PWE PB PROC ERP ERP-C-1210 0002 ASSISTANT EMERGENCY RESPONSE MANAGER (AERM) CANCELLED - REPLACED 10/24/95 PWE BY ERP-C-1200 PB PROC ERP ERP-C-1250 0004 EMERGENCY PREPAREDNESS COORDINATOR/EOF 06/25/01 PWE PB PROC ERP ERP-C-1250-1 0000 EMERGENCY POWER INSTRUCTIONS 09/14/94 PWE PB PROC ERP ERP-C-1250-2 0002 EMERGENCY PREPAREDNESS COORDINATOR INSTRUCTIONS FOR ASPEN 05/11/01 PWE BACKUP NOTIFICATION SYSTEM PB PROC ERP ERP-C-1250-3 0000 EMERGENCY PREPAREDNESS COORDINATOR INSTRUCTIONS TO STOP 09/14/94 PWE STAFFING PB PROC ERP ERP-C-1250-4 0000 EMERGENCY PREPAREDNESS COORDINATOR INSTUCTIONS FOR SYSTEM RESET 09/14/94 PWE PB PROC ERP ERP-C-1300 0010 EMERGENCY OPERATIONS FACILITY (EOF) DOSE ASSESSMENT TEAM LEADER 08/31/00 PWE PB PROC ERP ERP-C-1300-1 0004 DOSE ASSESSMENT TEAM LEADER (DATL) INITIAL ACTIONS 06/25/01 PWE PB PROC ERP ERP-C-1300-2 0000 DOSE ASSESSMENT TURNOVER LIST 09/23/94 PWE PB PROC ERP ERP-C-1300-3 0004 PROTECTIVE ACTION RECOMMENDATION WORKSHEET 03/30/01 PWE PB PROC ERP ERP-C-1300-4 0000 OFFSITE SAMPLE ANALYSIS REQUESTS 09/23/94 PWE PB PROC ERP ERP-C-1300-5 0001 DETERMINATION OF PROTECTIVE ACTION RECOMMENDATIONS (PARS) 11/02/98 PWE PB PROC ERP ERP-C-1300-6 0002 DOSE ASSESSMENT GROUP MEMBER (DAGM) INITIAL ACTIONS 06/25/01 PWE PB PROC ERP ERP-C-1300-7 0000 OBTAINING EPDS MET/RAD DATA 03/26/97 PWE PB PROC ERP ERP-C-1300-8 0000 USE OF MODE A/MODE B OF CDM 03/26/97 PWE PB PROC ERP ERP-C-1300-9 0001 OBTAINING MET DATA FROM NATIONAL WEATHER SERVICE 09/12/97 PWE PB PROC ERP ERP-C-1310 0003 EMERGENCY OPERATIONS FACILITY (EOF) DOSE ASSESSMENT GROUP - 03/26/97 PWE CANCELLED - REPLACED BY ERP-C-1300 PB PROC ERP ERP-C-1310-1 0000 DOSE ASSESSMENT GROUP LEADER INITIAL ACTIONS CANCELLED - 03/26/97 PWE REPLACED BY ERP-C-1300 PB PROC ERP ERP-C-1310-2 0000 OBTAINING MET DATA FROM NATIONAL WEATHER SERVICE CANCELLED - 03/24/97 PWE REPLACED BY ERP-C-1300 PB PROC ERP ERP-C-1310-3 0000 OBTAINING EPDS MET/RAD DATA - CANCELLED - NO REPLACED BY 03/26/97 PWE ERP-C-1300 PB PROC ERP ERP-C-1310-4 0000 USE OF MODE A/MODE B OF CDM CANCELLED - REPLACED BY ERP-C-1300 03/26/97 PWE PB PROC ERP ERP-C- 1320 0007 EMERGENCY OPERATIONS FACILITY (EOF) FIELD SURVEY GROUP LEADER 08/31/00 PWE PB PROC ERP ERP-C-1320-1 0002 FIELD SURVEY GROUP LEADER INITIAL ACTIONS 04/10/98 PWE PB PROC ERP ERP-C-1320-2 0001 FIELD SURVEY GROUP LEADER TURNOVER SHEET 03/26/97 PWE PB PROC ERP ERP-C-1320-3 0002 FIELD SURVEY GROUP LEADER DATA SHEET 08/31/00 PWE PB PROC ERP ERP-C-1400 0005 ENGINEERING SUPPORT TEAM 06/25/01 PWE PB PROC ERP ERP-C-1400-1 0002 ENGINEERING SUPPORT TEAM CHECKLIST 11/02/98 PWE PB PROC ERP ERP-C-1410 0002 CORE DAMAGE ASSESSMENT 09/09/98 PWE PB PROC ERP ERP-C-1410-1 0000 RADIOLOGICAL DATA 09/14/94 PWE PB PROC ERP ERP-C-1410-2 0001 HYDROGEN CONCENTRATION DATA 09/09/98 PWE

2/04/2002 PAGE 2 PEACH BOTTOM ATOMIC POWER STATION PROCEDURE INDEX REPORT:

CURR DOC PROC REV EFFECTIVE RESP SYSTEM FAC TYPE TYPE PROCEDURE NUMBER NBR TITLE DATE GROUP NBR PB PROC ERP ERP-C--1410-3 0001 CONTAINMENT RADIATION MONITOR DATA 09/09/98 PWE PB PROC ERP ERP-C-1410-4 0000 METAL WATER REACTION - CANCELLED NO REPLACEMENT 09/09/98 PWE PB PROC ERP ERP-C-1410-5 0002 PERCENT OF FUEL INVENTORY AIRBORNE IN THE CONTAINMENT 06/01/01 PWE VS. APPROXIMATE SOURCE AND DAMAGE ESTIMATE PB PROC ERP ERP-C-1410-6 0002 PROCEDURES FOR ESTIMATING FUEL DAMAGE BASED ON MEASURED 1-131 06/25/01 PWE AND XE-133 CONCENTRATIONS PB PROC ERP ERP-C-1500 0006 LOGISTICS SUPPORT TEAM 04/14/00 PWE PB PROC ERP ERP-C- 1500-1 0001 MESSAGE AND INFORMATION INSTRUCTIONS 10/24/95 PWE PB PROC ERP ERP-C-1500-2 0001 HELICOPTER LANDING INFORMATION 10/24/95 PWE PB PROC ERP ERP-C-1900 0004 RECOVERY PHASE IMPLEMENTATION 11/02/98 PWE PB PROC ERP ERP-C- 1900-1 0000 RECOVERY PHASE IMPLEMENTATION FLOW CHART 06/28/93 PWE PB PROC ERP ERP-C- 1900-2 0002 PEACH BOTTOM ATOMIC POWER STATION RECOVERY ACCEPTANCE CHECKLIST 04/02/98 PWE PB PROC ERP ERP-C- 1900-3 0002 LIMERICK GENERATING STATION RECOVERY ACCEPTANCE CHECKLIST 04/02/98 PWE PB PROC ERP ERP-C- 1900-4 0002 RECOVERY PLAN OUTLINE 04/02/98 PWE PB PROC ERP ERP-C- 1900-5 0002 ASSESSMENT CONSIDERATIONS 12/28/99 PWE PB PROC ERP ERP-101 0023 CLASSIFICATION OF EMERGENCIES 02/08/02 PWE PB PROC ERP ERP-101 BASES 0003 PBAPS EAL TECHNICAL BASIS MANUAL TABLE OF CONTENTS 02/08/02 PWE PB PROC ERP ERP-110 0013 EMERGENCY NOTIFICATIONS 05/11/01 PWE PB PROC ERP ERP-110 APP 1 0057 EMERGENCY NOTIFICATION TELEPHONE LIST 01/06/01 PWE PB PROC ERP ERP-110 APP 2 0024 EMERGENCY CLASSIFICATION NOTIFICATION TELEPHONE LIST FOR A SITE 07/21/93 PWE EMERGENCY OR GENERAL EMERGENCY CANCELLED - REPLACED BY ERP-110 APPENDIX 1 PB PROC ERP ERP-120 0002 PARTIAL PLANT EVACUATION CANCELLED - REPLACED BY ERP-130 & GP-15 08/10/92 PWE PB PROC ERP ERP-130 0018 SITE EVACUATION 11/20/01 PWE PB PROC ERP ERP-140 0019 EMERGENCY RESPONSE ORGANIZATION (ERO) CALL OUT 03/04/99 PWE PB PROC ERP ERP-140 APP 1 0019 AUTOMATED ERO ACTIVATION 08/06/98 PWE PB PROC ERP ERP-140 APP 2 0022 ASPEN EMERGENCY MESSAGE CANCELLED - REPLACED BY ERP-110 APP 1 08/06/98 PWE PB PROC ERP ERP-140 APP 3 0022 DOSE ASSESSMENT TEAM CANCELLED - REPLACED BY PIMS PRINTOUTS 08/20/92 ISSUED MONTHLY PER RT/ERP-2 PB PROC ERP ERP-140 APP 4 0015 CHEMISTRY SAMPLING & ANALYSIS TEAM CANCELLED - REPLACED BY PIMS 08/20/92 PRINTOUTS ISSUED MONTHLY PER RT/ERP-2 PB PROC ERP ERP-140 APP 5 0014 DAMAGE REPAIR TEAM CANCELLED - REPLACED BY PIMS PRINTOUTS ISSUED 08/20/92 MONTHLY PER RT/ERP-2 PB PROC ERP ERP-140 APP 6 0013 SECURITY TEAM CANCELLED - REPLACED BY PIMS PRINTOUTS ISSUED 08/20/92 MONTHLY PER RT/ERP-2 PB PROC ERP ERP-140 APP 7 0017 PERSONNEL SAFETY TEAM CANCELLED - REPLACED BY PIMS PRINTOUTS 08/20/92 ISSUED MONTHLY PER RT/ERP-2 PB PROC ERP ERP-140 APP 8 0009 COMPANY CONSULTANTS AND CONTRACTORS CANCELLED - INCLUDED IN 08/20/92 EMERGENCY TELEPHONE DIRECTORY PB PROC ERP ERP-140 APP 9 0011 NEARBY PUBLIC AND INDUSTRIAL USERS OF DOWNSTREAM WATER CANCELLED 08/20/92

- INCLUDED IN EMERGENCY TELEPHONE DIRECTORY PB PROC ERP ERP-200 0017 EMERGENCY DIRECTOR (ED) 03/27/01 PWE PB PROC ERP ERP-200 APP 1 0005 EMERGENCY DIRECTOR CHECKLIST (MCR) 12/10/01 PWE PB PROC ERP ERP-200 APP 2 0005 EMERGENCY DIRECTOR CHECKLIST (TSC) 03/30/01 PWE PB PROC ERP ERP-200 APP 3 0004 EVENT NOTIFICATION FORM 07/10/00 PWE PB PROC ERP ERP-200 APP 4 0004 STATION PUBLIC ADDRESS ANNONCEMENTS 07/10/00 PWE PB PROC ERP ERP-200 APP 5 0005 PAR DEVELOPMENT AND ISSUANCE 04/25/01 PWE PB PROC ERP ERP-200 APP 6 0001 DOSE ASSESSMENT DATA SHEET 07/10/00 PB PROC ERP ERP-200 APP 7 0000 TURNOVER/BREIFING FORM 07/10/00 PWE PB PROC ERP ERP-200 APP 8 0000 MINIMUM STAFFING POSITIONS NECESSARY TO ACTIVATE THE TSC 03/27/01 PWE

2/04/2002 PAGE 3 PEACH BOTTOM ATOMIC POWER STATION PROCEDURE INDEX REPORT:

CURR DOC PROC REV EFFECTIVE RESP SYSTEM FAC TYPE TYPE PROCEDURE NUMBER NBR TITLE DATE GROUP NBR PB PROC ERP ERP-205 0010 EMERGENCY PREPAREDNESS COORDINATOR/TSC 03/27/01 PWE PB PROC ERP ERP-206 0008 SUPPORT SERVICES GROUP 02/07/01 PWE PB PROC ERP ERP-210 0000 TRIP TABLE COMMUNICATOR (TSC) 09/12/97 PWE PB PROC ERP ERP-220 0006 OPERATIONS GROUP 10/05/95 PWE PB PROC ERP ERP-230 0016 OPERATIONS SUPPORT CENTER (OSC) ACTIVATION 10/07/98 PWE PB PROC ERP ERP-230 APP 1 0001 PERSONNEL EXPOSURE LOG OPERATIONS SUPPORT CENTER (OSC) 11/28/95 PWE CANCELLED - NO REPLACEMENT PB PROC ERP ERP-250 0011 TECHNICAL SUPPORT CENTER (TSC) ACTIVATION CANCELLED - NO 10/14/93 REPLACEMENT PB PROC ERP ERP-300 0007 DOSE ASSESSMENT TEAM LEADER (DATL) CANCELLED - NO REPLACEMENT 09/23/94 PWE PB PROC ERP ERP-301 0006 DOSE ASSESSMENT COORDINATOR (DAC) 04/25/01 PWE PB PROC ERP ERP-305 0004 DOSE ASSESSMENT GROUP LEADER (DAGL) CANCELLED - NO REPLACEMENT 03/12/93 PB PROC ERP ERP-306 0000 LIMERICK RESPONSE FOR SHIFT DOSE ASSESSMENT PERSONNEL (SDAP) 06/30/00 PWE PB PROC ERP ERP-310 0007 DOSE ASSESSMENT GROUP CANCELLED - NO REPLACEMENT 09/23/94 PWE PB PROC ERP ERP-315 0014 OPERATION OF THE DOSE ASSESSMENT COMPUTER 04/24/00 PWE PB PROC ERP ERP-318 0001 LIQUID RELEASE DOSE CALCULATIONS AT DOWNSTREAM WATER INTAKE 06/18/93 FACILITIES CANCELLED - REPLACED BY ERP-360 PB PROC ERP ERP-319 0001 LIQUID RELEASE DOSE CALCULATIONS FOR FISH INGESTION CANCELLED 06/18/93 REPLACED BY ERP-360 PB PROC ERP ERP-325 0005 SHIFT DOSE ASSESSMENT PERSONNEL 08/25/98 PWE PB PROC ERP ERP-325 APP 1 0000 CANCELLED - REPLACED BY MESOREM PROGRAM 03/03/95 PWE PB PROC ERP ERP-330 0009 FIELD SURVEY GROUP LEADER (FSGL) CANCELLED - NO REPLACEMENT 09/23/94 PWE PB PROC ERP ERP-340 0006 FIELD SURVEY GROUP 03/19/97 PWE PB PROC ERP ERP-340 APP 1 0005 FIELD SURVEY DATA SHEET 08/29/00 PWE PB PROC ERP ERP-360 0000 RADIOACTIVE LIQUID RELEASE CANCELLED - REPLACED BY ERP-315 06/23/94 PB PROC ERP ERP-400 0006 CHEMISTRY TEAM LEADER (CTL) 01/20/00 PWE PB PROC ERP ERP-410 0009 CHEMISTRY GROUP 04/30/98 PWE PB PROC ERP ERP-410 APP 1 0000 CHEMISTRY SAMPLE CHECK-OFF LIST CANCELLED - REPLACED BY 12/11/96 PWE ERP-410 PB PROC ERP ERP-410 APP 2 0000 CHEMISTRY SAMPLE AND ANALYSIS LOG SHEETCANCELLED - REPLACED BY 12/11/96 PWE ERP-410 PB PROC ERP ERP-500 0010 SECURITY TEAM LEADER (STL) 04/24/00 PWE PB PROC ERP ERP-510 0009 PERSONNEL ACCOUNTABILITY CANCELLED - NO REPLACEMENT 11/28/95 PWE PB PROC ERP ERP-520 0005 SECURITY GROUP LEADERS 11/28/95 PWE PB PROC ERP ERP-520 APP 1 0000 UNIT 1 PERSONNEL LOG CANCELLED - NO REPLACEMENT 11/28/95 PWE PB PROC ERP ERP-600 0013 HEALTH PHYSICS TEAM LEADER (HPTL) 07/07/99 PWE PB PROC ERP ERP-610 0004 FIRST AID/SEARCH AND RESCUE GROUP CANCELLED - NO REPLACEMENT 02/05/93 PB PROC ERP ERP-620 0012 HEALTH PHYSICS GROUP 10/13/00 PWE PB PROC ERP ERP-620 APP 1 0000 HABITABILITY STATUS LOG SHEET 11/05/93 PWE 101 PB PROC ERP ERP-620 APP 2 0000 ARM STATUS LOG 11/05/93 PWE 100 PB PROC ERP ERP-620 APP 3 0002 HEALTH PHYSICS BRIEFING GUIDE 09/04/98 PWE PB PROC ERP ERP-620 APP 4 0000 ACCESS BRIEFING GUIDE CANCELLED - NO REPLACEMENT 05/08/96 PWE PB PROC ERP ERP-630 0003 DOSIMETRY, BIOASSAY, AND RESPIRATORY PROTECTION GROUP CANCELLED 03/18/93

- NO REPLACEMENT PB PROC ERP ERP-640 0006 VEHICLE AND EVACUEE CONTROL GROUP 05/28/97 PWE PB PROC ERP ERP-640 APP 1 0000 CONTAMINATED VEHICLE SURVEY FORM CANCELLED - NO REPLACEMENT 05/28/97 PWE PB PROC ERP ERP-640 APP 2 0000 UNCONTAMINATED VEHICLE FORM CANCELLED - NO REPLACEMENT 05/28/97 PWE PB PROC ERP ERP-650 0006 TRANSPORT OF CONTAMINATED INJURY OFF-SITE 11/27/96 PWE PB PROC ERP ERP-660 0007 ENTRY FOR EMERGENCY REPAIR AND OPERATIONS CANCELLED - REPLACED 07/11/94 BY ERP-620

2/04/2002 PAGE 4 PEACH BOTTOM ATOMIC POWER STATION PROCEDURE INDEX REPORT:

CURR DOC PROC REV EFFECTIVE RESP SYSTEM FAC TYPE TYPE PROCEDURE NUMBER NBR TITLE DATE GROUP NBR PB PROC ERP ERP-670 0004 EMERGENCY RADIATION EXPOSURE GUIDELINES AND CONTROLS 12/11/96 PWE PB PROC ERP ERP-680 0007 CONTROL OF THYROID BLOCKING POTASSIUM IODIDE (KI) TABLETS 09/22/00 PWE PB PROC ERP ERP-680 APP 1 0001 POTASSIUM IODIDE WORKSHEET 02/20/97 PWE PB PROC ERP ERP-680 APP 2 0000 POTASSIUM IODIDE CONSENT FORM 11/30/94 PWE PB PROC ERP ERP-680 APP 3 0001 INSTRUCTION AND RECORD SHEET FOR PERSONS RECEIVING KI 02/20/97 PWE PB PROC ERP ERP-680 APP 4 0001 KI AUTHORIZATION 02/20/97 PWE PB PROC ERP ERP-700 0010 TECHNICAL SUPPORT TEAM 09/22/00 PWE PB PROC ERP ERP-710 0008 TECHNICAL SUPPORT GROUP CANCELLED - REPLACED BY ERP-700 11/02/98 PWE PB PROC ERP ERP-800 0006 OPERATIONS SUPPORT CENTER DIRECTOR (OSC DIRECTOR) 10/07/98 PWE PB PROC ERP ERP-810 0011 MAINTENANCE TEAM 07/07/99 PWE

• • END OF REPORT **