ML14164A073

Attachment 9: Discussion of Revision to the Radiological Emergency Plan Annex for Peach Bottom Atomic Power Station, Cover Through Page PBAPS 3-76

ML14164A073
Person / Time
Site:	Peach Bottom
Issue date:	05/30/2014
From:	Exelon Generation Co
To:	NRC/FSME, Office of Nuclear Material Safety and Safeguards
Shared Package
ML14164A053	List: ML14164A054 ML14164A055 ML14164A056 ML14164A058 ML14164A060 ML14164A061 ML14164A062 ML14164A063 ML14164A064 ML14164A065 ML14164A066 ML14164A068 ML14164A069 ML14164A071 ML14164A072 ML14164A073 ML14164A075 ML14164A076 ML14164A079 NEI 99-01, Braidwood, Units 1 and 2, Attachment 2, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1001, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 2A, and 2B, Cover Through Pa NEI 99-01, Byron, Units 1 and 2, Attachment 3, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1002, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosures 3A, 3B, Cover - by 3-74 NEI 99-01, Byron, Units 1 and 2, Attachment 3, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1002, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosures 3B, 3C, by 3-75 - by 3-157 NEI 99-01, Clinton, Unit 1, Attachment 4, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1003, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 4A and 4B, Cover - Cl 3-73 NEI 99-01, Clinton, Unit 1, Attachment 4, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1003, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 4B and 4C, Cl 3-74 - Cl 3-151 NEI 99-01, Dresden, Units 2 & 3, Attachment 5, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1004, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 5A and 5B, Cover - Dr 3-76 NEI 99-01, Dresden, Units 2 & 3, Attachment 5, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1004, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 5B and 5C, Dr 3-77 - Dr 3-160 NEI 99-01, LaSalle, Units 1 & 3, Attachment 6, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1005, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 6A and 6B, Cover - LS 3-74 NEI 99-01, License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. RA-14-032, Attachment 10: Discussion of Revision to the Radiological Emergency Plan Annex for Quad Cities Nuclear Power Station, Page QC 3-69 Through End RA-14-032, Attachment 11: Discussion of Revision to the Radiological Emergency Plan Annex for Three Mile Island Nuclear Station, Page TMI 3-80 Through End RA-14-032, Attachment 7: Discussion of Revision to the Radiological Emergency Plan Annex for Limerick Generating Station. Cover Through Page LGS 3-69 RA-14-032, Attachment 7: Discussion of Revision to the Radiological Emergency Plan Annex for Limerick Generating Station. LGS 3-70 Page LGS 3-155 RA-14-032, Attachment 8: Discussion of Revision to the Radiological Emergency Plan Annex for Oyster Creek Nuclear Generating Station, Page Ocgs 3-76 Through End RA-14-032, Attachment 8: Discussion of Revision to the Radiological Emergency Plan Annex for Oyster Creek Nuclear Generating Station. Cover Through Page Ocgs 3-75 RA-14-032, Attachment 9: Discussion of Revision to the Radiological Emergency Plan Annex for Peach Bottom Atomic Power Station, Cover Through Page PBAPS 3-76 RA-14-032, Attachment 9: Discussion of Revision to the Radiological Emergency Plan Annex for Peach Bottom Atomic Power Station, Page PBAPS 3-77 Through End RA-14-032, Braidwood, Units 1 and 2, Attachment 2, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1001, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 2A, and 2B, Cover Through Pa RA-14-032, Byron, Units 1 and 2, Attachment 3, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1002, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosures 3A, 3B, Cover - by 3-74 RA-14-032, Byron, Units 1 and 2, Attachment 3, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1002, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosures 3B, 3C, by 3-75 - by 3-157 RA-14-032, Clinton, Unit 1, Attachment 4, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1003, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 4A and 4B, Cover - Cl 3-73 RA-14-032, Clinton, Unit 1, Attachment 4, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1003, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 4B and 4C, Cl 3-74 - Cl 3-151 RA-14-032, Dresden, Units 2 & 3, Attachment 5, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1004, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 5A and 5B, Cover - Dr 3-76 RA-14-032, Dresden, Units 2 & 3, Attachment 5, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1004, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 5B and 5C, Dr 3-77 - Dr 3-160 RA-14-032, LaSalle, Units 1 & 3, Attachment 6, Discussion of Revision to the Radiological Emergency Plan Annex, EP-AA-1005, NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, Enclosure 6A and 6B, Cover - LS 3-74 RA-14-032, License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. RS-14-115, Attachment 6: Discussion of Revision to the Radiological Emergency Plan Annex for LaSalle County Station. LS 3-75 Through LS 3-150 TMI-14-046, Attachment 8: Discussion of Revision to the Radiological Emergency Plan Annex for Oyster Creek Nuclear Generating Station, Page Ocgs 3-76 Through End
References
RA-14-032, RS-14-115, TMI-14-046 EP-AA-1007, NEI 99-01, Rev. 6
Download: ML14164A073 (144)

Text

ATTACHMENT 9DISCUSSION OF REVISIONTO THERADIOLOGICAL EMERGENCY PLAN ANNEXFORPEACH BOTTOM ATOMIC POWER STATIONEP-AA-1007 Enclosures

• Enclosure 9A -EAL Comparison Matrix Document* Enclosure 9B -EAL Red-Line Basis Document* Enclosure 9C -EAL Basis Document NEI 99-01REVISION 6DEVELOPMENT OF EMERGENCY ACTIONLEVELS FOR NON-PASSIVE REACTORSATTACHMENT 9DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FORPEACH Bo'nOM ATOMIC POWER STATIONExelon Generation.

0SEAL Comoarison Matrix DocumentNEI 99-01 Rev 6 1 Proposed EAL I Justification AG1Initiating Condition

-GENERAL EMERGENCY Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mRemTEDE or 5,000 mRem thyroid CDE.Operating Mode Applicability:

AllExample Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the General Emergency promptlyupon determining that the applicable time has been exceeded, or will likely beexceeded.

• If an ongoing release is detected and the release start time is unknown,assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due toactions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.

• The pre-calculated effluent monitor values presented in EAL #1 should beused for emergency classification assessments until the results from a doseassessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than thereading shown for 15 minutes or longer:(site-specific monitor list and threshold values)2. Dose assessment actual meteorology indicates doses greater than 1000mRem TEDE or 5000 mRem thyroid CDE at or beyond (site-specific dosereceptor point)3. Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):, Closed window dose rates greater than 1000 mR/hr expected tocontinue for 60 minutes or longer.* Analysis of field survey samples indicate thyroid CDE greater than5000 mRem for one hour of inhalation.

RG1Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mRemTEDE or 5,000 mRem thyroid CDE.Operating Mode Applicability:

1,2,3,4,5,D Emergency Action Level (EAL):Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

• If an ongoing release is detected and the release start time is unknown,assume that the release duration has exceeded 15 minutes.* Classification based on effluent monitor readings assumes that a release pathto the environment is established.

If the effluent flow past an effluent monitor isknown to have stopped due to actions to isolate the release path, then theeffluent monitor reading is no longer valid for classification purposes.

• The pre-calculated effluent monitor values presented in EAL #1 should beused for emergency classification assessments until the results from a doseassessment using actual meteorology are available.

1. Readings on ANY Table RI Effluent Monitor > Table R1 value for > 15minutes.OR2. Dose assessment using actual meteorology indicates doses at or beyond thesite boundary of EITHER:a. >1000mRemTEDE ORb. > 5000 mRem CDE ThyroidOR3. Field survey results at or beyond the site boundary indicate EITHER:a. Gamma (closed window) dose rates >1000 mRihr are expected tocontinue for > 60 minutes.ORb. Analyses of field survey samples indicate

> 5000 mRem CDEThyroid for 60 minutes of inhalation.

Table R1 Effluent Monitor Thresholds Release Path General Emergency Main Stack (RI-0-1 7-0508 Common) 5.57 E+09 uCi/secVent Stack (RI-2979B Unit 2 or 1.60 E+08 uCi/secRI-397/B Unit 3)I No Change EI] Difference Deviation

1) Listed site-specific monitors and Threshold values to ensure timelyclassification.

2) Added the following to bullet #3" Classification based on effluent monitorreadings assumes that a release path to the environment is established."

In orderto delete the following from the basis "Classification based on effluent monitorreadings assumes that a release path to the environment is established.

If theeffluent flow past an effluent monitor is known to have stopped due to actions toisolate the release path, then the effluent monitor reading is no longer valid forclassification purposes."

This allows for more timely classification since all thebasis information pertaining to Note bullet 3 will be contained in the IC and thereforreadily available on the 1 1x17 procedure matrix used by the SM.Page 1 of66 00EAL Comparison Matrix DocumentNEI 99-01 Rev 6 1 Proposed EAL Justification ASIInitiating Condition

-SITE AREA EMERGENCY Release of gaseous radioactivity resulting in offsite dose greater than 100 mRemTEDE or 500 mRem thyroid CDE.Operating Mode Applicability:

AllExample Emergency Action Levels: (1 or 2 or 3)Notes:The Emergency Director should declare the Site Area Emergency promptlyupon determining that the applicable time has been exceeded, or will likely beexceeded.

If an ongoing release is detected and the release start time is unknown,assume that the release duration has exceeded 15 minutes.If the effluent flow past an effluent monitor is known to have stopped due toactions to isolate the release path, then the effluent monitor reading is nolonger valid for classification purposes.

The pre-calculated effluent monitor values presented in EAL #1 should beused for emergency classification assessments until the results from a doseassessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than thereading shown for 15 minutes or longer:(site-specific monitor list and threshold values)2. Dose assessment actual meteorology indicates doses greater than 1000mRem TEDE or 5000 mRem thyroid CDE at or beyond (site-specific dosereceptor point)3. Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):* Closed window dose rates greater than 100 mR/hr expected tocontinue for 60 minutes or longer.* Analysis of field survey samples indicate thyroid CDE greater than500 mRem for one hour of inhalation.

RS1Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mRemTEDE or 500 mRem thyroid CDE.Operating Mode Applicability:

1,2,3.4,5,D Emergency Action Level (EAL):Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

" If an ongoing release is detected and the release start time is unknown,assume that the release duration has exceeded 15 minutes.* Classification based on effluent monitor readings assumes that a release pathto the environment is established.

If the effluent flow past an effluent monitor isknown to have stopped due to actions to isolate the release path, then theeffluent monitor reading is no longer valid for classification purposes.

" The pre-calculated effluent monitor values presented in EAL #1 should be usedfor emergency classification assessments until the results from a doseassessment using actual meteorology are available.

1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15minutes.OR2. Dose assessment using actual meteorology indicates doses at or beyond thesite boundary of EITHER:a. > 100 mRem TEDEORb. > 500 mRem CDE ThyroidOR3. Field survey results at or beyond the site boundary indicate EITHER:a. Gamma (closed window) dose rates >100 mR/hr are expected tocontinue for > 60 minutes.ORb. Analyses of field survey samples indicate

> 500 mRem CDEThyroid for 60 minutes of inhalation.

Table RI Effluent Monitor Thresholds

-No Change E Difference 1 Deviation

1) Listed site-specific monitors and Threshold values to ensure timelyclassification.

2) Added the following to bullet #3" Classification based on effluent monitorreadings assumes that a release path to the environment is established.!'

In orderto delete the following from the basis "Classification based on effluent monitorreadings assumes that a release path to the environment is established.

If theeffluent flow past an effluent monitor is known to have stopped due to actions toisolate the release path, then the effluent monitor reading is no longer valid forclassification purposes."

This allows for more timely classification since all thebasis information pertaining to Note bullet 3 will be contained in the IC and thereforreadily available on the 11x17 procedure matrix used by the SM.Release PathMain Stack (RI-0-17-050/B Common)Vent Stack (RI-2979B Unit 2 orRI-3971B Unit 3)Site Area Emer gency5.57 E+OU ucL/sec1.60 E+07 uCi/secPage 2 of 66 EAL Comnarison Matrix DocumentNEI 99-01 Rey 6 Proposed EAL Justification AA1Initiating Condition

-ALERTRelease of gaseous or liquid radioactivity resulting in offsite dose greater than 10mRem TEDE or 50 mRem thyroid CDE.Operating Mode Applicability:

AllExample Emergency Action Levels: (1 or 2 or 3)Note:* The Emergency Director should declare the Alert promptly upondetermining that the applicable time has been exceeded, or will likely beexceeded.

• If an ongoing release is detected and the release start time is unknown.assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped dueto actions to isolate the release path, then the effluent monitor reading isno longer valid for classification purposes.

• The pre-calculated effluent monitor values presented in EAL #1 should beused for emergency classification assessments until the results from adose assessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than thereading shown for 15 minutes or longer:(site-specific monitor list and threshold values)2. Dose assessment actual meteorology indicates doses greater than 10mRem TEDE or 50 mRem thyroid CDE at or beyond (site-specific dosereceptor point)3. Analysis of a liquid effluent sample indicates a concentration or releaserate that would result in doses greater than 10 mRem TEDE or 50 mRemthyroid CDE at or beyond (site-specific dose receptor point) for one hour ofexposure.

4. Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):* Closed window dose rates greater than 10 mR/hr expected tocontinue for 60 minutes or longer.* Analysis of field survey samples indicate thyroid CDE greater than50 mRem for one hour of inhalation.

RA1Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10mRem TEDE or 50 mRem thyroid CDE.Operating Mode Applicability:

1,2, 3,4, 5, DEmergency Action Level (EAL):Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

• If an ongoing release is detected and the release start time is unknown, assumethat the release duration has exceeded 15 minutes.* Classification based on effluent monitor readings assumes that a release pathto the environment is established.

If the effluent flow past an effluent monitor isknown to have stopped due to actions to isolate the release path, then theeffluent monitor reading is no longer valid for classification purposes.

• The pre-calculated effluent monitor values presented in EAL #1 should be usedfor emergency classification assessments until the results from a doseassessment using actual meteorology are available.

1. Readings on ANY Table R1 Effluent Monitor > Table RI value for> 15minutes.OR2. Dose assessment using actual meteorology indicates doses at or beyond thesite boundary of EITHER:a. > 10 mRem TEDEORb. > 50 mRem CDE ThyroidOR3. Analysis of a liquid effluent sample indicates a concentration or release ratethat would result in doses greater than EITHER of the following at or beyondthe site boundarya. 10 mRem TEDE for 60 minutes of exposureORb. 50 mRem CDE Thyroid for 60 minutes of exposureOR4. Field survey results at or beyond the site boundary indicate EITHER:a. Gamma (closed window) dose rates > 10 mRlhr areexpected to continue for 1 60 minutes.ORb. Analyses of field survey samples indicate

> 50 mRemCDE Thyroid for 60 minutes of inhalation.

Table R1 Effluent Monitor Thresholds Release Path AlertMain Stack (RI-0-17-050B Common) 5.57 E+07 uCi/secVent Stack (RI-2979B Unit 2 or 1.60 E+06 uCi/secRI-3979B Unit 3)No Change [ -] Difference

--1 Deviation

1) Listed site-specific monitors and Threshold values to ensure timely classification.

2) Added the following to bullet #3" Classification based on effluent monitorreadings assumes that a release path to the environment is established."

In orderto delete the following from the basis "Classification based on effluent monitorreadings assumes that a release path to the environment is established.

If theeffluent flow past an effluent monitor is known to have stopped due to actions toisolate the release path, then the effluent monitor reading is no longer valid forclassification purposes."

This allows for more timely classification since all the basisinformation pertaining to Note bullet 3 will be contained in the IC and thereforreadily available on the 1 1x17 procedure matrix used by the SM.3) A calculation was performed, in accordance with (lAW) guidance provided in NEI99-01 revision 6 EAL AA1, to determine the effluent monitor response for aradioactive liquid release via the normal site release pathway.

The liquid releasewould contain activity equivalent to provide 10mrem TEDE or 50mrem thyroid CDEat the site boundary.

The calculation determined the effluent monitor responsewould be >110% of the instrument maximum range and as such, lAW NEI 99-01Rev 6 guidance, was not included in this EAL.Page 3 of 66 0EAL Comoarison Matrix DocumentEAL Connnarison Matrix DocumentNEI 99-01 Rev 6Proposed EALJustification AU1Initiating Condition

-UNUSUAL EVENTRelease of gaseous or liquid radioactivity greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes or longerOperating Mode Applicability:

AllExample Emergency Action Levels: (1 or 2 or 3)Note:RUIInitiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for60 minutes or longer.Operating Mode Applicability:

1,2, 3, 4, 5, DEmergency Action Level (EAL):Notes:D: No Change E Difference 1-1 Deviation

1) Listed site-specific monitors and Threshold values to ensure timely classification.

2) Added the following to bullet #3" Classification based on effluent monitorreadings assumes that a release path to the environment is established."

In orderto delete the following from the basis "Classification based on effluent monitorreadings assumes that a release path to the environment is established.

If theeffluent flow past an effluent monitor is known to have stopped due to actions toisolate the release path, then the effluent monitor reading is no longer valid forclassification purposes."

This allows for more timely classification since all the basisinformation pertaining to Note bullet 3 will be contained in the IC and thereforreadily available on the 1 1x17 procedure matrix used by the SM.The Emergency Director should declare the Unusual Event promptly upon

• The Emergency Director should declare the event promptly upon determining determining that 60 minutes has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded.

exceeu .* If an ongoing release is detected and the release start time is unknown,assume that the release duration has exceeded 60 minutes." If the effluent flow past an effluent monitor is known to have stopped,indicating that the release path is isolated, the effluent monitor reading isno longer valid for classification purposes.

1. Reading on ANY effluent radiation monitor greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes orlonger:(site-specific monitor list and threshold values corresponding to 2 timesthe controlling document limits)2. Reading on ANY effluent radiation monitor greater than 2 times thealarm setpoint established by a current radioactivity discharge permit for60 minutes or longer.3. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times (site-specific effluent releasecontrolling document limits) for 60 minutes or longer.* If an ongoing release is detected and the release start time is unknown, assumethat the release duration has exceeded 60 minutes.Classification based on effluent monitor readings assumes that a release pathto the environment is established.

If the effluent flow past an effluent monitor isknown to have stopped due to actions to isolate the release path, then theeffluent monitor reading is no longer valid for classification purposes.

1. Reading on ANY of the following effluent monitors

> 2 times alarmsetpoint established by a current radioactive release discharge permitfor k 60 minutes." Radwaste Discharge Effluent Monitor (RI-0-1 7-350)OR" Discharge Permit specified monitorOR2. Readings on ANY Table RI Effluent Monitor > Table R1 value for > 60minutes:Table R1 Effluent Monitor Thresholds Release Path Unusual EventMain Stack(RI-O-17-050B Common) 6.36 E+05 uCi/secMain Stack(RI-O-17-050A Common) 6.36 E+05 uCi/secVent Stack (RI-2979B Unit 2 or 1.82 E+04 uCi/secRI-3979B Unit 3)1.82 E+04 uCi/secVent Stack (RI-2979A Unit 2 orRI-3979A Unit 3)OR3. Confirmed sample analyses for gaseous or liquid releases indicateconcentrations or release rates > 2 times ODCM Limit with a releaseduration of > 60 minutes.Page 4 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification AG2 RG2Initiating Condition

-GENERAL EMERGENCY A---- No Change F -] Difference Deviation Spent fuel pool level cannot be restored to at least (site-specific Level 3description) for 60 minutes or longer. 1) EAL not used in accordance with the discussion in Section 1.4, NRC OrderEA-1 2-051. It is recommended that this EAL be implemented when the enhancedOperating Mode Applicability:

All spent fuel pool level instrumentation is available for use. The completion of theenhanced SFP level indicators and need for the inclusion of this EAL is beingExample Emergency Action Levels: tracked in accordance with Exelon Generation

Company, LLC's Initial StatusReport to March 12, 2012 Commission Order Modifying Licenses with Regard forNOTES: The Emergency Director should declare the General Emergency Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) datedpromptly upon determining that 60 minutes has been exceeded, October 25,2012.or will likely be exceeded1. Spent fuel pool level cannot be restored to at least (site-specific Level 3description) for 60 minutes or longer.Page 5 of 66 0 0 0EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification AS2 RS2Initiating Condition

-SITE AREA EMERGENCY

[ No Change I Difference Deviation Spent fuel pool level cannot be restored to at least (site-specific Level 3description)

1) EAL not used in accordance with the discussion in Section 1.4, NRC OrderEA-1 2-051. It is recommended that this EAL be implemented when the enhancedOperating Mode Applicability:

All spent fuel pool level instrumentation is available for use. The completion of theenhanced SFP level indicators and need for the inclusion of this EAL is beingExample Emergency Action Levels: tracked in accordance with Exelon Generation

Company, LLC's Initial StatusReport to March 12, 2012 Commission Order Modifying Licenses with Regard for1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dateddescription)

October 25,2012.Page 6 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition

-ALERT AA2 Initiating Condition:

RA2 No Change Difference Deviation Significant lowering of water level above, or damage to, irradiated fuel. Significant lowering of water level above, or damage to, irradiated fuel. 1) Listed site-specific monitors and Threshold values to ensure timely classification.

Operating Mode Applicability:

All Operating Mode Applicability:

1, 2,3,4,5, D 2) EAL #3 not used in accordance with the discussion in Section 1.4, NRC OrderEA-1 2-051. It is recommended that this EAL be implemented when the enhancedExample Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): spent fuel pool level instrumentation is available for use. The completion of theenhanced SFP level indicators and need for the inclusion of this EAL is being1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

tracked in accordance with Exelon Generation

Company, LLC's Initial StatusOR Report to March 12, 2012 Commission Order Modifying Licenses with Regard for2. Damage to irradiated fuel resulting in a release of radioactivity from the Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) datedfuel as indicated by ANY of the following radiation monitors:

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel October 25,2012.as indicated by ANY Table R2 Radiation Monitor reading >1000 mRemlhr.(site-specific listing of radiation

monitors, and the associated

readings, setpoints and/or alarms) Table R23. Lowering of spent fuel pool level to (site-specific Level 2 value). Refuel Floor ARM's" Steam Separator Pool, 234 ELEV, 3.7 (7.9)" Refuel Slot, 234 ELEV, 3.8 (7.10)" Fuel Pool, 234 ELEV, 3.9 (7.11)* Refueling Bridge, 234 ELEV. 3.10 (7.12)Page 7 of 66 0EAL Comparison Matrix DocumentNEI 99-01 Rev 6 1 Proposed EAL Justification AU2Initiating Condition:

UNUSUAL EVENTUNPLANNED loss of water level above irradiated fuelOperating Mode Applicability:

AllExample Emergency Action Levels:1. a. UNPLANNED water level drop in the REFUELING PATHWAY asindicated by ANY of the following:

(site-specific level indications).

ANDb. UNPLANNED rise in area radiation levels as indicated by ANY of thefollowing radiation monitors.

(site-specific list of area radiation monitors)

RU2Initiating Condition:

UNPLANNED loss of water level above irradiated fuelOperating Mode Applicability:

1,2, 3,4,5, DEmergency Action Level (EAL):1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated byANY of the following:

" Refueling Cavity water level < 458 inches on Shutdown/Refuel Range indicator LI-2(3)-2-3-86 or PR/FR-2(3)-06-097.

OR* Spent Fuel Pool level < 232 feet 3 inches plant elevation.

OR" Indication or report of a drop in water level in the REFUELING PATHWAY.ANDb. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors inTable R2.E No Change 1 Difference

--- Deviation

1) Listed site-specific level indication and monitors to ensure timelyclassification.

Table R2Refuel Floor ARM's* Steam Separator Pool, 234 ELEV, 3.7 (7.9)" Refuel Slot, 234 ELEV, 3.8 (7.10)" Fuel Pool, 234 ELEV, 3.9 (7.11)" Refueling Bridge, 234 ELEV, 3.10 (7.12)Page 8 of 66 EAL Comnari~on Matrix DocumentNEI 99-01 Rev 6 Proposed EAL I Justification AA3Initiating Condition

-ALERTRadiation levels that impede access to equipment necessary for normal plantoperations, cooldown or shutdown.

Operating Mode Applicability:

AllExample Emergency Action Levels: (1 or 2)Note: If the equipment in the listed room or area was already inoperable, orout of service, before the event occurred, then no emergency classification is warranted

1. Dose rate greater than 15 mR/hr in ANY of the following areas:* Control Room* Central Alarm Station* (other site-specific areas/rooms)

2. An UNPLANNED event results in radiation levels that prevent orsignificantly impede access to any of the following plant rooms orareas:(site-specific list of plant rooms or areas with entry-related modeapplicability identified)

RA3Initiating Condition:

Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

Operating Mode Applicability:

1,2, 3, 4, 5,DEmergency Action Level (EAL):Note:No Change 1 Difference 1 Deviation

1) Listed site-specific plant rooms and areas with identified mode applicability to ensure timely classification.

• If the equipment in the room or area listed in Table R3 was already inoperable, orout of service, before the event occurred, then no emergency classification iswarranted.

1. Dose rate> 15 mR/hr in ANY of the following Table R3 areas:Table R3Areas Reauirina Continuous Occuoancv I* Main Control Room* Central Alarm Station -(by survey)OR2. UNPLANNED event results in radiation levels that prohibit or significantly impedeaccess to ANY of the following Table R4 plant rooms or areas:Table R4Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building*

Modes 3, 4 and 5*Areas required to establish shutdown coolingPage 9 of 66 EAL Comparison Matrix DocumentNEt 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENT Initiating Condition:

RU3 [ No Change H Difference Deviation Reactor coolant activity greater than Technical Specification allowable limits. Reactor coolant activity greater than Technical Specification allowable limits. 1) Listed site-specific monitor and Threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

2) Listed this system category EAL in the radiological category EAL section toPower Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3 maintain consistency with current and previous revisions of Exelon EALs. This willExample Emergency Action Levels: Emergency Action Level (EAL): ensure a timely classification since the threshold values are more aligned with theradiological category vice system category.

1. (Site-specific radiation monitor) reading greater than (site-specific value). 1. Air Ejector discharge radiation monitor RIS-2(3)-17-150 A/B Hi-HI alarm.OR OR2. Sample analysis indicates that a reactor coolant activity value is greater than 2. Specific coolant activity

> 4.0 uClIgm Dose equivalent 1-131.an allowable limit specified in Technical Specifications.

Page 10 of 66 EAL Comparison Matrix DocumentNEI 99-01 rev 6 Fission Product Barrier MatrixFisso Product __BarrierMatrixHotIMatrix GEEA EMREC SIT ARE EMREC LRFG1 Loss of any two barriers AND Loss or Potential Loss of third barrier.1,2,3FSI Loss or Potential Loss of ANY two barriers.

1,2,3FA1 ANY Loss or ANY Potential Loss of either Fuel Clad or RCS1,2,3FC -Fuel Clad RC -Reactor Coolant System CT -Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss1. RCS Activity I A. UNPLANNED rapid drop in primary A. Primary containment pressure greaterPrimary containment pressure following than (site-specific value)Containment A. Primary containment pressure primary containment pressure rise ORreactor coolant activity is greater NoneA.Piaycnimetpsur Pressure

/ greater than (site-specific value) None OR B. (site-specific explosive mixture) existsPrimary than 300 uCi/gm dose equivalent due to RCS leakage.

B. Primary containment pressure inside primary containment Containment 1-131) response not consistent with LOCA ORConditions conditions.

C. HCTL exceeded.

A. RPV water level cannot be restored A. RPV water level cannot beA. Primary containment flooding and maintained above (site-specific restored and maintained above2. RPV Water required.

RPV water level corresponding to (site-specific RPV water level None None A. Primary containment flooding required.

Level top of active fuel) or cannot be corresponding to the top of activedetermined.

fuel) or cannot be determined.

A. UNISOLABLE direct downstream pathway to the environment existsafter primary containment isolation signalORA. UNISOLABLE primary system leakageA. UNISOLABLE break in any of the that results in exceeding EITHER of B. Intentional primary containment 3.RCS Leak Rate/ following:

( site-specific systems with the following.

venting per EOPsPrimary None None potential for high-energy line breaks) 1.ax Nomal Operating O NoneContainment OR Temperature C. UNISOLABLE primary systemIsolation Failure B. Emergency RPV Depressurzation OR leakage that results in exceeding

2. Max Normal Operating Area EITHER of the following:

Radiation Level. 1. Max Safe Operating Temperature.

OR2. Max Safe Operating AreaRadiation Level.4.Primary A. Primary Containment Radiation A. Primary Containment Radiation A. Primary Containment Radiation MonitorContainment Monitor reading greater than None Monitor reading greater than (site- None None reading greater than (site-specific Radiation (site-specific value), specific value). value).A. Any Condition in the opinion of A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A Any Condition in the opinion of the A. Any Condition in the opinion of the5. Emergency the Emergency Director that Emergency Director that indicates Emergency Director that indicates Loss Emergency Director the iniofte A.rAny Diton inat indicates Loss Emergency Director that indicates Potential Director Judgment indicates Loss of the Fuel Clad Potential Loss of the Fuel Clad Barrier.

of the RCS Barrier.

Potential Loss of the RCS BarEer. of the Containment Barrier.

Loss of the Containment Baenier.Barrier.Page 11 of 66 EAL Comparison Matrix DocumentProposed Fission Product Barrier MatrixI_____________

IEEAEMREC SITE_______

_ ARE EMERGENCY ALERT HotMatriFGI Loss of any two barriers AND Loss or Potential Loss of third barrier,1 FSI Loss or Potential Loss of ANY two barriers.

FAI ANY Loss or ANY Potential Loss of either Fuel Clad or RCSFC -Fuel Clad RC -Reactor Coolant System CT -Containment Loss Poetential Less Less Petentla Loee Loes Potential LossCoolant activity s 300 uCIggm Dose None None1. RCS Activity Equivalent 1-131. None None None2. RPV water level cannot be restored and 1. RPV water level cannos be restored and2 RPV Water 1. Plant conditions indicate Primary maintained

> -172 tIches (TAF) maintained

> -172 Inches (TAF) Plant condities indicate Primary Centainment Level Containment flooding is required.

OR OR None None fleodisg is required.

3. RPV water level cannot be determined.

2. RPV water level cannot be deternined.

3. Primary Containment pressure

> e6 psig.OR1 UNPLANNED rapid drop in Drywell 4. a. Drywell or toms hydrogen concentration I. Drywell pressure

>2.0 pstg. pressure following primary containment

> 6%.3. Primary AND pressure rise. ANDContainment None None 2 Drywell pressure rise is due to RCS None OR b. Drywell or toms oxygen concentration Pressure/Conditions leakage 2 Drywell pressure response sot consistent

> 5%.with LOCA conditions.

OR5. Heat Capacity Limit (T-102 Curve T/T-t)exceeded.

1. UNISOLABLE Main Steam Line (MSL), 3. UNISOLABLE primary system leakage thatHPCI, Feedwater, RWCU or RCIC line results in EITHER of the following:

break, a. Secondary Containment area4.RCS Leak Rate None None OR temperature

' T-103 Alarm Setpolnt None None2. Emergency RPV Depressurization is ORrequired.

b. Secondary Containment area radiation level > T-103 Alarm Setpotst.

5.Primary Drywell radiation monitor reading Drywell radiation monitor reading Drywell radiation monitor readingContainment

> 9.55 E+02 R/hr. None > lgORfhr.

None None > 2.20 E+O3 R/hrRadiation

1. UNISOLABLE direct downstream pathway tothe environment exists after primarycontainment isolation signal.OR2. Intentional Primary Containment venting/purging per EOPs or $AMPs due to6eprimary accident conditions.

Containment None None None None OR NoneIsolation Failure 3. UNISOLABLE primary system leakage thatresults in EITHER of the following:

a. Secondary Containment areatemperature

> T-103 Action Level.ORb. Secondary Containment arearadiation level > T-IO3 Action Le el.Any Condition in the opinion of the Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the Emergency Emergency Emergency Director that indicates Loss Emergency Director that indicates Potential Emergency Director that indicates Loss of Emergency Director that indicates Potential Emergency Director that indicates Less of the Director that indicates Potential Loss of theDirector Judgment of the Fuel Clad Barrier.

Loss of the Fuel Clad Barrier, the RCS Barrier Loss of the RCS Barrier.

Containment Barner. Containment Banier.Page 12 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Fuel Clad Barrier FC Category:

Fuel Clad BarrierFC No Change Diffrence Deviation RCS Activity RCS Activity

1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2,3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. (Site-specific indications that reactor coolant activity is greater than 300 uCi/gm Coolant activity

> 300 uCi/gm Dose Equivalent 1-131.dose equivalent 1-131)Page 13 of66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Fuel Clad Barrier FC2 Category:

Fuel Clad BarrierFC2 W N Chg Differe Deviation RPV Water Level RCS Activity

1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2,3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. Primary containment flooding required.

1. Plant conditions indicate Primary Containment flooding is required.

Potential Loss Potential LossA. RPV water level cannot be restored and maintained above (site-specific RPV 2. RPV water level cannot be restored and maintained

> -172 Inches (TAF)water level corresponding to top of active fuel) or cannot be determined.

OR3. RPV water level cannot be determined.

Page 14 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Fuel Clad Barrier FC4 Category:

Fuel Clad Barrier FC5 E No Change Difference Deviation Primary Containment Radiation Primary Containment Radiation

1) Listed site-specific monitor and threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LosSA. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 9.55 E+02 R/hr.value).Page 15 of'66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Fuel Clad Barrier FC6 Category:

Fuel Clad Barrier FC7 No Change Difference Deviation Emergency Director Judgment Emergency Director JudgmentOperating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. Any Condition in the opinion of the Emergency Director that indicates Loss ofthe Fuel Clad Barrier.

the Fuel Clad Barrier.Potential Loss Potential LossA. Any Condition in the opinion of the Emergency Director that indicates Potential

2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.

Loss of the Fuel Clad Barrier.Page 16 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Reactor Coolant System Barrier Rd Category:

Reactor Coolant System Barrier RC3 [ No Change Difference Deviation Primary Containment Pressure Primary Containment Pressure/Conditions

1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1, 2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. Primary containment pressure greater than (site-specific value) due to RCS 1. Drywell pressure

>2.0 psig.leakage.

AND2. Drywell pressure rise is due to RCS leakage.Page 17 of'66 EAL Comnarison Matrix DocumentEAL Cornnarison Matrix DocumentNEI 99-01 Rev 6Proposed EALJustification 4 I.RC2Category:

Reactor Coolant System BarrierRPV Water LevelOperating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownFission Product Barrier Threshold:

LossA. RPV water level cannot be restored and maintained above (site-specific RPVlevel corresponding to the top of active fuel) or cannot be determined.

RC2Category:

Reactor Coolant System BarrierRPV Water LevelOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LosS1. RPV water level cannot be restored and maintained

> -172 inches (TAF).OR2. RPV water level cannot be determined.

E No Change [: Difference 1:1 Deviation 1 ) Listed site-specific threshold value to ensure timely classification.

Page 18 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Reactor Coolant System Barrier RC3 Category:

Reactor Coolant System Barrier RC4 E No Change Difference Deviation RCS Leak Rate RCS Leak Rate 1) Listed site-specific systems and threshold values to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. UNISOLABLE break in any of the following:

( site-specific systems with potential

1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC linefor high-energy line breaks) break.OR ORB. Emergency RPV Depressurization

2. Emergency RPV Depressurization is required.

Potential Loss Potential LossA. UNISOLABLE primary system leakage that results in exceeding EITHER of thefollowing:

3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature

> T-103 Alarm Setpoint.

1. Max Normal Operating Temperature.

OROR b. Secondary Containment area radiation level > T-103 Alarm Setpoint.

2. Max Normal Operating Area Radiation Level.Page 19 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Reactor Coolant System Barrier RC4 Category:

Reactor Coolant System BarrierRC No Change Diff n Deviation Primary Containment Radiation Primary Containment Radiation

1) Listed site-specific monitor and threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 100 R/hr.value).Page 20 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Reactor Coolant System Barrier RC6 Category:

Reactor Coolant System Barrier RC7 No Change Diffrence Deviation Emergency Director Judgment Emergency director JudgmentOperating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. ANY Condition in the opinion of the Emergency Director that indicates Loss ofthe RCS Barrier.

the RCS Barrier.Potential Loss Potential LossA. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.

2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.Page 21 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Containment Barrier Category:

Containment Barrier C No Change Difference Deviation Primary Containment Conditions Primary Containment Pressure/Conditions

1) Listed site-specific threshold values to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

2) The words "and rising" were added to account for the momentary spike inPower Operation,

Startup, Hot Standby, Hot Shutdown

1. 2, 3 pressure where pressure is now lowering, the risk of a potential loss ofcontainment is no longer present, this wording is also consistent with present EALFission Product Barrier Threshold:

Fission Product Barrier (FPB) Threahold:

wording.Loss LossC. UNPLANNED rapid drop in primary containment pressure following primary 1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.containment pressure rise OROR 2. Drywell pressure response not consistent with LOCA conditions.

B. Primary containment pressure response not consistent with LOCA conditions.

Potential LossPotential Loss0. Primary containment pressure greater than (site-specific value) 3. Drywell pressure

> 56 psig and rising.OROR 4. a. Drywell or Torus Hydrogen concentration

> 6%.E. (site-specific explosive mixture) exists inside primary containment ANDOR b. Drywell or Torus Oxygen concentration

> 5%.OR3. HCTL exceeded.

5. Heat Capacity Limit (T-102 Curve T/T-l) exceeded.

Page 22 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CT, CT2 D N hne D Dfeec eitoCategory:

Containment Barrier Category:

Containment Barrier No Change Difference l Deviation RPV Water Level RPV Water LevelOperating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2,3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Potential Loss Potential LossA. Primary containment flooding required.

Plant conditions indicate Primary Containment flooding is required.

Page 23 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Containment Barrier CT3 Category:

Containment Barrier CT6 W No Change Difference Deviation Primary Containment Isolation Failure Primary Containment Isolation Failure 1) Listed site-specific threshold values to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1, 2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. UNISOLABLE direct downstream pathway to the environment exists after 1. UNISOLABLE direct downstream pathway to the environment exists afterprimary containment isolation signal primary containment isolation signal.OR OR2. Intentional Primary Containment venting/purging per EOPs or SAMPs due toB. Intentional primary containment venting per EOPs accident conditions.

OR ORC. UNISOLABLE primary system leakage that results in exceeding EITHER of the 3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature

> T-103 Action Levelfollowing:

OR1. Max Safe Operating Temperature.

b. Secondary Containment area radiation level > T-103 Action LevelOR2. Max Safe Operating Area Radiation Level.Page 24 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Category:

Containment Barrier CT4 Category:

Containment Barrier CT E No Change Diffrence Deviation Primary Containment Radiation Primary Containment Radiation

1) Listed site-specific monitor and threshold value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2, 3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Potential Loss Potential LossA. Primary Containment Radiation Monitor reading greater than (site-specific value). Drywell radiation monitor reading > 2.20 E+03 R/h.Page 25 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CT6 CT7 [ ohne [ Dfeec ]eitoCategory:

Containment Barrier Category:

No Change Difference Deviation Emergency director Judgment Emergency Director JudgmentOperating Mode Applicability:

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot Shutdown 1,2,3Fission Product Barrier Threshold:

Fission Product Barrier (FPB) Threshold:

Loss LossA. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. Any Condition in the opinion of the Emergency Director that indicates Loss ofthe Containment Barrier.

the Containment Barrier.Potential Loss Potential LossA. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.

2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.Page 26 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL I Justification SG1Initiating Condition:

GENERAL EMERGENCY Prolonged loss of all offsite and all onsite AC power to emergency buses.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the General Emergency promptlyupon determining that (site-specific hours) has been exceeded, or willlikely be exceeded.

1. a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).ANDb. EITHER of the following:

" Restoration of at least one emergency bus in less than(site-specific hours) is not likely." (Site-specific indication of an inability to adequately remove heat from the core)MG11Initiating Condition:

Prolonged loss of all offsite and all onsite AC power to emergency buses.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1 Loss of ALL offsite AC power to unit 4 KV Safeguards Buses.AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply power tounit 4KV Safeguards Buses.AND3. EITHER of the following:

a. Restoration of at least one unit 4KV safeguards Bus in < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is notlikely.ORb. RPV water level cannot be restored and maintained

> -195 inches.No Change FIDifference I Deviation

1) Listed site-specific equipment, site-specific time based on station blackoutcoping analysis, and site-specific indication to ensure timely classification.

Page 27 of 66 EAL Comoarison Matrix DocumentNEI 99-01 Rev 6SS1Initiating Condition:

SITE AREA EMERGENCY Loss of all offsite and all onsite AC power to emergency buses for 15 minutes orlonger.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, orwill likely be exceeded.

Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses)for 15 minutes or longer.Proposed EAL Justification MS1Initiating Condition:

Loss of all offsite and onsite AC power to emergency buses for 15 minutes orlonger.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC Power to unit 4KV Safeguards Buses.AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply power tounit 4KV Safeguards Buses.AND3. Failure to restore power to at least one unit 4KV Safeguards bus in < 15minutes from the time of loss of both offsite and onsite AC power.No Change FIDifference FII Deviation

1) Listed site-specific equipment to ensure timely classification.

Page 28 of 66 EAL Comnarican Matrix Oo~zimnnt NEI 99-01 Rev 6 Proposed EAL Justification SA1IInitiating Condition:

ALERTLoss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.

1. a. AC power capability to (site-specific emergency buses) is reduced to asingle power source for 15 minutes or longer.ANDb. Any additional single power source failure will result in loss of all ACpower to SAFETY SYSTEMS.MAIInitiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. AC power capability to unit 4KV Safeguards Buses reduced to only one of thefollowing power sources for > 15 minutes.* 2 Emergency Auxiliary Transformer (OAX04)* 3 Emergency Auxiliary Transformer (OBX04)* El Diesel Generator

• E2 Diesel Generator

• E3 Diesel Generator

• E4 Diesel Generator AND2. ANY additional single power source failure will result in a loss of ALL AC powerto SAFETY SYSTEMS.M No Change --Difference FIDeviation

1) Listed site-specific equipment to ensure timely classification.

Page 29 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 1 Proposed EAL Justification SulInitiating Condition:

UNUSUAL EVENTLoss of all offsite AC power capability to emergency buses for 15 minutes orlonger.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the Unusual Event promptly upondetermining that 15 minutes has been exceeded, or will likely beexceeded.

Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15minutes or longerMU1Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes orlonger.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

Loss of ALL offsite AC power capability to unit 4KV Safeguards Buses for> 15 minutes.M No Change R Difference FIDeviation

1) Listed site-specific equipment to ensure timely classification.

Page 30 of 66 EAL Comnarison Matrix DoumimntNEI 99-01 Rev 6 1 Proposed EAL I Justification SGSInitiating Condition:

GENERAL EMERGENCY Loss of all AC and Vital DC power sources for 15 minutes or longer.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the General Emergency promptlyupon determining that 15 minutes has been exceeded, or will likely beexceeded.

1 Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses) for 15 minutes or longer.ANDIndicated voltage is less than (site-specific bus voltage value) on ALL(site-specific vital DC buses) for 15 minutes or longer.MG2Initiating Condition:

Loss of all AC and Vital DC power sources for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to unit 4KV safeguards Buses.AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply powerto unit 4KV Safeguards Buses.AND3. Voltage is < 107.5 VDC on unit 125 VDC battery buses 2(3)0D021, 2(3)0D022, 2(3)0D023, 2(3)0D024.

AND4. ALL AC and Vital DC power sources have been lost for > 16 minutes.D No Change F Difference F -Deviation

1) Listed site-specific equipment to ensure timely classification.

2) Removed the word "indicated",

this will allow for an indication problem to notcause confusion on the need to declare.Page 31 of 66 EAL Comoarison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification SS5Initiating Condition:

SITE AREA EMERGENCY Loss of all Vital DC power for 15 minutes or longer.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, orwill likely be exceeded.

Indicated voltage is less than (site-specific bus voltage value) on ALL Vital DCbuses for 15 minutes or longer.MS2Initiating Condition:

Loss of all Vital DC power for 15 minutes or longer.Operating Mode Applicability:

1,2,3D No Change M Difference FIDeviation

1) Listed site-specific equipment and site-specific value to ensure timelyclassification.

2) Removed the word "indicated",

this will allow for an indication problem to notcause confusion on the need to declare.Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

Voltage is <107.5 VDC on 125 VDC battery buses 2(3)0D021, 2(3)0D022, 2(3)0D023, 2(3)0D024 for > 15 minutes.Page 32 of 66 EAL Comoarison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification SS5MS3Initiating Condition:

SITE AREA EMERGENCY Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPVwater level [BWR]) or RCS heat removal.Operating Mode Applicability:

Power Operation Example Emergency Action Levels:1 .a. An automatic (trip [PWR] / scram [BWR]) did not shutdown thereactor.ANDb. All manual actions to shutdown the reactor have been unsuccessful.

ANDc. EITHER of the following conditions exist:1. (Site-specific indication of an inability to adequately remove heat from the core)OR2. (Site-specific indication of an inability to adequately remove heat from the RCS)Initiating Condition:

Inability to shutdown the reactor causing a challenge to RPV water level or RCSheat removal.Operating Mode Applicability:

1,2Emergency Action Level (EAL):1. Automatic scram did not shutdown the reactor as indicated by Reactor Power> 4%.AND2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful asindicated by Reactor Power > 4%.AND3. EITHER of the following conditions exist:" RPV water level cannot be restored and maintained

> -195 inchesOR" Heat Capacity Limit (T-102 Curve T/T-1) exceeded.

D- No Change Difference FIDeviation

1) Listed site-specific indications to ensure timely classification.

2) Mode 2 included in operating mode applicability as per developer notes.3) Added ARI as an equivalent scram.Page 33 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification SA5Initiating Condition:

ALERTAutomatic or manual (trip [PWR] I scram [BWR]) fails to shutdown the reactor, andsubsequent manual actions taken at the reactor control consoles are notsuccessful in shutting down the reactor.Operating Mode Applicability:

Power Operation Example Emergency Action Levels:Note: A manual action is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core, and does not include manuallydriving in control rods or implementation of boron injection strategies.

I. a. An automatic (trip [PWR I scram [BWRI) did not shutdown thereactor.ANDb. Manual action taken at the reactor control consoles are notsuccessful in shutting down the reactor.MA3Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manualactions taken at the reactor control consoles are not successful in shutting downthe reactor.Operating Mode Applicability:

1,2Emergency Action Level (EAL):Note:* A manual action is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core, and does not include manuallydriving in control rods or implementation of boron injection strategies.

1. Automatic or manual scram did not shutdown the reactor as indicated byReactor Power > 4%.AND2. Manual / ARI actions taken at the Reactor Console are not successful inshutting down the reactor as indicated by Reactor Power > 4%.D No Change -q Difference FIDeviation

1) Listed site-specific indications to ensure timely classification.

2) Mode 2 included in operating mode applicability as per developer notes.3) Added ARI as an equivalent scram.Page 34 of 66 EAL Comoarison Matrix DocumentNEI 99-01 Rev 6 1 Proposed EAL Justification SU*Initiating Condition:

UNUSUAL EVENTAutomatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor.Operating Mode Applicability:

Power Operation Example Emergency Action Levels: (1 or 2)Note: A manual action is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core, and does not include manuallydriving in control rods or implementation of boron injection strategies.

1. a. An automatic (trip [PWR] I scram [BWR]) did not shutdown thereactor.ANDb. A subsequent manual action taken at the reactor control consoles issuccessful in shutting down the reactor.2. a. A manual scram ([PWR] / scram [BWR]) did not shutdown thereactor.ANDb. EITHER of the following:

1. A subsequent manual action taken at the reactor control consolesis successful in shutting down the reactor.OR2. A subsequent automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor.MU3Initiating Condition:

Automatic or manual scram fails to shutdown the reactor.Operating Mode Applicability:

F No Change M Difference FIDeviation

1) Listed site-specific indications to ensure timely classification.

2) Mode 2 included in operating mode applicability as per developer notes.3) Added ARI as an equivalent scram.1,2Emergency Action Level (EAL):Note:A manual action is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core, and does not include manuallydriving in control rods or implementation of boron injection strategies.

a. Automatic scram did not shutdown the reactor as indicated by ReactorPower > 4%.ANDb. Subsequent manual / ARI action taken at the Reactor Console issuccessful in shutting down the reactor.OR2. a. Manual scram did not shutdown the reactor as indicated by ReactorPower > 4%.ANDb. EITHER of the following:

1. Subsequent manual / ARI action taken at the Reactor Console issuccessful in shutting down the reactor.OR2. Subsequent automatic/

ARI scram is successful in shutting downthe reactor.Page 35 of 66 0EAL Comoarison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL I Justification SA2MA4Initiating Condition:

ALERTUNPLANNED loss of Control Room indications for 15 minutes or longer with asignificant transient in progress.

Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the Alert promptly upondetermining that 15 minutes has been exceeded, or will likely beexceeded.

1. a. An UNPLANNED event results in the inability to monitor one or more of thefollowing parameters from within the Control Room for 15 minutes or longer.[see table below][ BWR parameter list] [PWR parameter list]M No Change FI Difference FIDeviation Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with asignificant transient in progress.

Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. UNPLANNED event results in the inability to monitor ANY Table M1parameters from within the Control Room for >15 minutes.Reactor Power Reactor PowerRPV Level RCS LevelRPV Pressure RCS PressurePrimary Containment Pressure In Core/Core Exit Temperature Levels in at least (site-specific Suppression Pool Level number) steam generators Suppression Pool Temperature Steam Generator Auxiliary orEmergency Feed Water FlowTable Ml Control Room Parameters

• Reactor Power* RPV Water Level* RPV Pressure* Primary Containment Pressure* Torus Level* Torus Temperature AND 12. ANY Table M2 transient in progress.

ANDb. Any of the following transient events in progress.

• Automatic or Manual runback greater than 25% thermal reactor power* Electrical load rejection greater than 25% full electrical load* Reactor Scram [BWR] / trip [PWR]* ECCS (SI) actuation

• Thermal power oscillations greater than 10% [BWR]Table M2 Significant Transients

• Automatic or Manual Runback >25% thermal reactor power* Electrical Load Rejection

>25% full electrical load* Reactor Scram* ECCS Actuation

• Thermal Power oscillations

> 10%Page 36 of 66 EAL ComDarison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification SU2Initiating Condition:

UNUSUAL EVENTUNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:Note: The Emergency Director should declare the Unusual Event promptly upondetermining that 15 minutes has been exceeded, or will likely beexceeded.

An UNPLANNED event results in the inability to monitor one or more of thefollowing parameters from within the Control Room for 15 minutes or longer.[see table below][ BWR parameter list] [PWR parameter list]Reactor Power Reactor PowerRPV Level RCS LevelRPV Pressure RCS PressurePrimary Containment Pressure In Core/Core Exit Temperature Levels in at least (site-specific Suppression Pool Level number) steam generators Steam Generator Auxiliary orSuppression Pool Temperature Emergency Feed Water FlowMU4Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

UNPLANNED event results in the inability to monitor ANY Table M1 parameters from within the Control Room for > 15 minutes.Table M1 Control Room Parameters

" Reactor Power* RPV Water Level" RPV Pressure* Primary Containment Pressure* Torus LevelTorus Temperature No Change FIDifference FIDeviation Page 37 of 66 0EAL Comparison Matrix Docuiment NEI 99-01 Rev 6 Proposed EAL Justification SA9Initiating Condition:

ALERTHazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels:1. a. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

• Internal or external flooding event* High winds or tornado strike* FIRE* EXPLOSION

• (site-specific hazards)* Other events with similar hazard characteristics asdetermined by the Shift ManagerANDb. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for thecurrent operating mode.OR2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEMcomponent or structure needed for the current operating mode.MA5Initiating Condition:

ALERTHazardous event affecting a SAFETY SYSTEM required for the current operating mode.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):1. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

• Internal or external flooding event* High winds or tornado strike" FIRE" EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift ManagerAND2. EITHER of the following:

a. Event damage has caused indications of degradedperformance in at least one train of a SAFETY SYSTEMrequired by Technical Specifications for the currentoperating mode.ORb. The event has caused VISIBLE DAMAGE to a SAFETYSYSTEM component or structure required by Technical Specifications for the current operating mode.D No Change M Difference F Deviation

1) No additional site-specific hazards noted2) Changed the word "needed" to "required" in the IC and "required by Technical Specifications" in the EAL to be consistent with terminology used by operators andminimize confusion.

Page 38 of 66 EAL Comnarison Matrix DocumentNEI 99-01 Rev 6SU4Initiating Condition:

UNUSUAL EVENTRCS leakage for 15 minutes or longer.Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the Unusual Event promptly upondetermining that 15 minutes has been exceeded, or will likely beexceeded.

1. RCS unidentified or pressure boundary leakage greater than (site-specific value) for 15 minutes or longer.2. RCS identified leakage greater than (site-specific value) for 15 minutes orlonger3. Leakage from the RCS to a location outside containment greater than 25gpm for 15 minutes or longerProposed EAL I Justification MU6Initiating Condition:

RCS leakage for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for> 15 minutes.OR2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.OR3. Leakage from the RCS to a location outside the Drywell >25 gpm for > 15minutes.[I. No Change -- Difference

-IDeviation

1) Listed site-specific values to ensure timely classification.

2) In EAL #3 Changed wording from containment to Drywell for clarity to betterdefine the primary containment structure.

3) In EAL #1 and 2, added "into the Drywell" to differentiate between EAL #1/2 and#3. Without this wording would have been in EAL #1 or #2 concurrent with #3. Withthe added wording each EAL can be called separately.

Page 39 of 66 CAl f, M.*.. fl.- .,NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENTLoss of all onsite or offsite communications capabilities Operating Mode Applicability:

Power Operation,

Startup, Hot Standby, Hot ShutdownExample Emergency Action Levels: (1 or 2 or 3)1. Loss of ALL of the following onsite communication methods:(site-specific list of communications method2. Loss of ALL of the following ORO communications s) methods:(site-specific list of communications methods)3. Loss of ALL of the following NRC communications methods:(site-specific list of communications methods)IUBMU7Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

1,2,3M No Change F ]Difference 1 Deviation

1) Listed site-specific communication methods to ensure timely classification.

Emergency Action Level (EAL):1. Loss of ALL Table M3 Onsite communications capability affecting theability to perform routine operations.

OR2. Loss of ALL Table M3 Offisite communication capability affecting theability to perform offsite notifications.

OR3. Loss of ALL Table M3 NRC communication capability affecting the abilityto perform NRC notifications.

Table M3 Communications Capability System Onsite Offsite NRCStation Radio System XPlant Public Address (PA) XOMNI System X XStation Phones X X XSatellite Phones X X XNARS XHPN X XENS X XPage 40 of 66 EAL ComDarison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL [ Justification CA2Initiating Condition:

ALERTLoss of all offsite and all onsite AC power to emergency buses for 15 minutes orlonger.Operating Mode Applicability:

Cold Shutdown, Refueling, DefueledExample Emergency Action Levels:Note: The Emergency Director should declare the Alert promptly upondetermining that 15 minutes time has been exceeded, or will likely beexceeded.

Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses)for 15 minutes or longer.CAIInitiating Condition:

Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability:

4,5, DEmergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining thatthe applicable time has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to unit 4KV Safeguards Buses.AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply powerto unit 4KV Safeguards Buses.AND3. Failure to restore power to at least one unit 4KV Safeguards bus in< 15 minutes from the time of loss of both offsite and onsite AC power.No Change 1 Difference 1-1 Deviation

1) Listed site-specific equipment to ensure timely classification.

Page 41 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENT CU2 Initiating Condition:

Cull No Change Difference Deviation Loss of all but one AC power source to emergency buses for 15 minutes or longer. Loss of all but one AC power source to emergency buses for 15 minutes or 1) Listed site-specific equipment to ensure timely classification.

Operating Mode Applicability:

longer.Cold Shutdown, Refueling, Defueled Operating Mode Applicability:

Example Emergency Action Levels: 4, 5, DNote: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):determining that 15 minutes time has been exceeded, or will likely be exceeded.

Note:1. a. AC power capability to (site-specific emergency buses) is reduced to a The Emergency Director should declare the event promptly upon determining single power source for 15 minutes or longer, that the applicable time has been exceeded, or will likely be exceeded.

AND 1. AC power capability to unit 4KV Safeguards Buses reduced to only oneb. Any additional single power source failure will result in loss of all AC of the following power sources for > 15 minutes.power to SAFETY SYSTEMS.

-2 Emergency Auxiliary Transformer (OAX04)* 3 Emergency Auxiliary Transformer (0BX04)* El Diesel Generator

• E2 Diesel Generator

• E3 Diesel Generator

• E4 Diesel Generator AND2. ANY additional single power source failure will result in a loss of ALL ACpower to SAFETY SYSTEMS.Page 42 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 [ Proposed EAL Justification CA6Initiating Condition

-ALERTHazardous event affecting SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels:1. a. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

• Internal or external flooding event* High winds or tornado strike* FIRE" EXPLOSION

• (site-specific hazards)* Other events with similar hazard characteristics asdetermined by the Shift ManagerANDb. EITHER of the following:

1. Event damage has caused indications of degradedperformance in at least one train of a SAFETY SYSTEMneeded for the current operating mode.OR2. The event has caused VISIBLE DAMAGE to a SAFETYSYSTEM component or structure needed for the currentoperating mode.CA2Initiating Condition:

Hazardous event affecting SAFETY SYSTEM required for the current operating mode.Operating Mode Applicability:

4,5Emergency Action Level (EAL):1. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

• Internal or external flooding event* High winds or tornado strike* FIRE* EXPLOSION

• Other events with similar hazard characteristics asdetermined by the Shift ManagerAND2. EITHER of the following:

a. Event damage has caused indications of degradedperformance in at least one train of a SAFETY SYSTEMrequired by Technical Specifications for the currentoperating mode.ORb. The event has caused VISIBLE DAMAGE to a SAFETYSYSTEM component or structure required by Technical Specifications for the current operating mode.H: No Change E Difference Deviation

1) No additional site-specific hazards noted2) Changed the word "needed" to "required" in the IC and "required byTechnical Specifications" in the EAL to be consistent with terminology used byoperators and minimize confusion.

Page 43 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CU4 CU3 N hne [ ifrne eitoInitiating Condition:

UNUSUAL EVENT Initiating Condition:

CU1 No Change Difference Deviation Loss of Vital DC power for 15 minutes or longer. Loss of Vital DC power for 15 minutes or longer. 1) Listed site-specific voltage and equipment to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

2) Removed the word "indicated",

this will allow for an indication problem to notCold Shutdown, Refueling 4, 5 cause confusion on the need to declare.Example Emergency Action Levels: Emergency Action Level (EAL):Note: The Emergency Director should declare the Unusual Event promptly upon Note:determining that 15 minutes time has been exceeded, or will likely be The Emergency Director should declare the event promptly upon determining

exceeded, that the applicable time has been exceeded, or will likely be exceeded.

Indicated voltage is less than (site-specific bus voltage value) on required Vital DCbuses for 15 minutes or longer.Voltage is < 107.5 VDC on required 125 VDC battery buses 2(3)0D021, 2(3)0D022, 2(3)0D023, and 2(3)0D024 for > 15 minutes.Page 44 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL I Justification CU5Initiating Condition:

UNUSUAL EVENTLoss of all onsite or offsite communications capabilities Operating Mode Applicability:

Cold Shutdown, Refueling, DefuledExample Emergency Action Levels: (1 or 2 or 3)1. Loss of ALL of the following onsite communication methods:(site-specific list of communications method2. Loss of ALL of the following ORO communications s) methods:(site-specific list of communications methods)3. Loss of ALL of the following NRC communications methods:(site-specific list of communications methods)CU4Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

4,5, DEmergency Action Level (EAL):1. Loss of ALL Table C1 Onsite communications capability affecting theability to perform routine operations.

OR2. Loss of ALL Table Cl Offite communication capability affecting theability to perform offsite notifications.

OR3. Loss of ALL Table C1 NRC communication capability affecting theability to perform NRC notifications.

Table C1 Communictions Capaeility System Onsite Offaite NRCStation Radio System XPlant Public Address (PA) XOMNI System X XStation Phones X X XSatellite Phones X X XNARS XHPN X XENS X XE No Change D Difference 1:1 Deviation

1) Listed site-specific communications methods to ensure timely classification.

Page 45 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

ALERT CA3 Initiating Condition:

CAS W No Change Difference Deviation Inability to maintain the plant in cold shutdown.

Inability to maintain plant in cold shutdown.

1) Listed site-specific Technical Specifications cold shutdown temperature limit andOperating Mods Applicability:

Operating Mode Applicability:

site-specific pressure reading to ensure timely classification.

Cold Shutdown, Refueling 4, 5Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):Note: The Emergency Director should declare the Alert promptly upon Note:determining that the applicable has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining exceeded.

that the applicable time has been exceeded, or will likely be exceeded.

1. UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown temperature limit) for greater thanthe duration specified in the following table. 1. UNPLANNED rise in RCS temperature>

21r2F due to loss of decay2. UNPLANNED RCS pressure increase greater than (site-specific pressure heat removal for > Table C2 duration.

reading).

(This EAL does not apply during water-solid plant conditions.

[PWR]) Table C2 RCS Heat-up Duration Thresholds Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up DurationRCS Status Containment Closure Heat-up Duration StatusStatus Intact Not Applicable 60 minutes*Intact (but not Established 20 minutes*RCS Reduced Not Applicable 60 minutes*

Not IntactInventory

[PWR])Not Intact (or at Established 20 minutes*

Not Established 0 minutesreduced inventory

[PWR]) Not Established 0 minutes

• If an RCS heat removal system is in operation within this timeframe and RCS temperature is being reduced, then EAL #1 is no._t* If an RCS heat removal system is in operation within this time frame and applicable.

RCS temperature is being reduced, the EAL is not applicable.

OR2. UNPLANNED RPV pressure rise > 10 psig as a result of temperature rise due to loss of decay heat removal.Page 46 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CU3 CU5 m m D eeiInitiating Condition:

UNUSUAL EVENT Initiating Condition:

[-J No Change Difference Deviation UNPLANNED increase in RCS temperature.

UNPLANNED rise in RCS temperature.

1) Listed site-specific Technical Specifications cold shutdown temperature limitOperating Mode Applicability:

Operating Mode Applicability:

to ensure timely classification.

Cold Shutdown, Refueling 4, 5 2) Changed the word increase to rise in the initiating condition to be consistent Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL): with operations language and training.

Note: The Emergency Director should declare the Unusual Event promptly Note:upon determining that 15 minutes time has been exceeded, or will likelybe exceeded.

The Emergency Director should declare the event promptly upondetermining that the applicable time has been exceeded, or will likely be1. UNPLANNED increase in RCS temperature to greater than (site-specific exceeded.

Technical Specification cold shutdown temperature limit).2. Loss of ALL RCS temperature and (reactor vessel/RCS

[PWR] or RPV 1. UNPLANNED rise in RCS temperature>

212F due to loss of decay[BWR]) level indication for 15 minutes or longer. heat removal.OR2. Loss of the following for >1 5 minutes.* ALL RCS temperature indications AND* ALL RPV water level indications Page 47 of 66 CAI t, RH f ; r%_ _ _'-__ _ _-.I. pc. .CJI *n at. IA r .x .I ,IcU lne,NEI 99-01 Rev 6 Proposed EAL Justification CGIInitiating Condition:

GENERAL EMERGENCY Loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory affecting fuel cladintegrity with containment challenged.

Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes time has been exceeded, orwill likely be exceeded.

1. a. (Reactor vessel/RCS

[PIVR] or RPV [BWrR]) vessel level less than (site-specific level) for 30 minutes or longer.ANDb. ANY indication from the Containment Challenge Table2. a.. (Reactor vessel/RCS

[PWR] or RPV [BLVR]) vessel level cannot bemonitored for 30 minutes or longer.ANDb. Core uncovery is indicated by ANY of the following:

• (Site-specific radiation monitor) reading greater than (site-specific value)* Erratic source range monitor indication

[PAR]* UNPLANNED increase in (site-specific sump and/or tank levels) ofsufficient magnitude to indicate core uncovery* (Other site-specific indications)

ANDc. ANY indication from the Containment Challenge Table).CG6Initiating Condition:

Loss of RPV inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

4,5Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining that theapplicable time has been exceeded, or will likely be exceeded.

1. a. RPV water level < -172 Inches (TAF) for > 30 mtnutes.ANDb. ANY Containment Challenge Indication (Table C4)OR2. a. RPV water level unknown for > 30 minutes.ANDb. Core uncovery is indicated by ANY of the following:

Table C3 indications of a sufficient magnitude to indicate core uncovery.

ORANY Table C5 Refuel Floor Area Radiation Monitor >3 Rihr.ANDc. ANY Containment Challenne Indication (Table C4)1H No Change E Difference 1 Deviation

1) Listed site-specific levels, radiation

monitors, and sumps and tanks to ensuretimely classification.

2) Listed Explosive mixture in the Containment Challenge Table to ensuretimely classification.

3) Worded "cannot be monitored" as unknown to ensure clarity for instances when the indicator is working, but is over/under ranged. This is also in keepingwith current EAL wording.Table C3 Indications of RCS LeakageUNPLANNED floor or equipment sump level risedUNPLANNED Tows level riseoUNPLANNED vessel make up rate riseObservation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

Table: Containment Challenge Table* CONTAINMENT CLOSURE not established*

• (Explosive mixture) exists inside containment

• UNPLANNED increase in containment pressure" Secondary containment radiation monitor reading above (site-specific value) [BWR]if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency isnot required.

Table C4 Containment Challenge Indications

• Primary Containment Hydrogen Concentration

> 6% and Oxygen> 5%UNPLANNED rise in containment pressureCONTAINMENT CLOSURE not established*

ANY Secondary Containment radiation monitor > T-103 Action Level* if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncoverytime limit, then escalation to a General Emergency is Mot required.

Table cs7Table C5Refuel Floor ARM'sSteam Separator Pool, 234 ELEV, 3.7 (7.9)Refuel Slot, 234 ELEV, 3.8 17.10)Fuel Pool, 234 ELEV, 3.9 (7.11)Refueling Bridge, 234 ELEV, 3.10 (7.12)__ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _I _ _ _ -_ _ _ _ _ _ _ _ _ _ _ _ -_ _ _ I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _Page 48 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CS1Initiating Condition:

SITE AREA EMERGENCY Loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory affecting coredecay heat removal capability.

Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes time has been exceeded, orwill likely be exceeded.

1. a. CONTAINMENT CLOSURE not established.

ANDb. (Reactor vessel/RCS

[PlR] or RPV [BW4RI) level less than (site-specific level).2. a. CONTAINMENT CLOSURE established.

ANDb. (Reactor vessel/RCS

[PWR] or RPV [BWlR) level less than (site-specific level).3. a. (Reactor vessel/RCS

[PWR] or RPV [BWR]) level cannot bemonitored for 30 minutes or longer.ANDb. Core uncovery is indicated by ANY of the following:

" (Site-specific radiation monitor) reading greater than (site-specific value)" Erratic source range monitor indication

[PWR]" UNPLANNED increase in (site-specific sump and/or tanklevels) of sufficient magnitude to indicate core uncovery" (Other site-specific indications)

CS6Initiating Condition:

Loss of RPV inventory affecting core decay heat removal capabilities.

Operating Mode Applicability:

4,5Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining thatthe applicable time has been exceeded, or will likely be exceeded.

1. With CONTAINMENT CLOSURE not established.

RPV water level < -160InchesOR2. With CONTAINMENT CLOSURE established, RPV water level < -172 Inches(TAF)OR3. a. RPV water level unknown for > 30 minutesANDb. Core uncovery is indicated by ANY of the following:

• Table C3 indications of a sufficient magnitude to indicate coreuncovery.

OR* ANY Table C5 Refuel Floor Area Radiation Monitor >3 R/hrD No Change E Difference

[j] Deviation

1) Listed site-specific values for level, radiation

monitors, and sumps and tanksto ensure timely classification.

2) Worded "cannot be monitored" as unknown to ensure clarity for instances when the indicator is working, but is over/under ranged. This is also in keepingwith current EAL wording.Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise*UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise* Observation of leakage or inventory lossRise in level is attributed to a loss of RPV inventory.

Table C6Refuel Floor ARM's* Steam Separator Pool, 234 ELEV, 3.7(7.9)* Refuel Slot, 234 ELEV, 3.8 (7.10)* Fuel Pool, 234 ELEV, 3.9 (7.11)* Refueling Bridge, 234 ELEV, 3.10 (7.12)Page 49 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CAIInitiating Condition:

ALERTLoss of (reactor vessel/RCS

[PWVR] or RPV [BWR]) inventory Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Alert promptly upondetermining that 15 minutes time has been exceeded, or will likely beexceeded.

1. Loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory asindicated by level less than (site-specific level).2. a. (Reactor vessel/RCS

[PWR] or RPV [BWRI) level cannot bemonitored for 15 minutes or longerANDb. UNPLANNED increase in (site-specific sump and/or tank) levels dueto a loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory.

CA6Initiating Condition:

Loss of RPV inventory Operating Mode Applicability:

4,5Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upondetermining that the applicable time has been exceeded, or will likely beexceeded.

1. Loss of RPV inventory as indicated by level < -48 inches.OR2. a. RPV water level unknown for> 15 minutes.ANDb. Loss of RPV inventory per Table C3 indications.

D No Change W Difference Deviation

1) Listed site-specific levels, and sumps and tanks to ensure timelyclassification.

2) Worded "cannot be monitored" as unknown to ensure clarity for instances when the indicator is working, but is over/under ranged. This is also in keepingwith current EAL wording.Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise* Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

Page 50 of 66 EAL Comoarison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification CulInitiating Condition:

UNUSUAL EVENTUNPLANNED loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory for15 minutes or longer.Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Unusual Event promptly upondetermining that 15 minutes has been exceeded, or will likely beexceeded.

1. UNPLANNED loss of reactor coolant results in (reactor vessel/RCS

[PVVR] or RPV [BWR]) level less than a required lower limit for 15minutes or longer.2. a. (Reactor vessel/RCS

[PWR] or RPV [BlWR) level cannot bemonitored.

ANDb. UNPLANNED increase in (site-specific sump and/or tank) levels.CU6Initiating Condition:

UNPLANNED loss of RPV inventory for 15 minutes or longer.Operating Mode Applicability:

4,5Emergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upondetermining that the applicable time has been exceeded, or will likely beexceeded.

1. UNPLANNED loss of reactor coolant results in the inability to restore andmaintain RPV water level above the procedurally established lower limitfor> 15 minutes.OR2. a. RPV water level unknownANDb. Loss of RPV inventory per Table C3 indications.

D No Change E Difference 1 Deviation

1) Described "a required lower limit" as a procedurally established lower limit,and listed site-specific sumps and tanks to ensure timely classification.

2) Worded "cannot be monitored" as unknown to ensure clarity for instances when the indicator is working, but is over/under ranged. This is also in keepingwith current EAL wording.Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise* Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

Page 51 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification HG1 HG1Initiating Condition:

GENERAL EMERGENCY Initiating Condition:

1 No Change Difference Deviation HOSTILE ACTION resulting in loss of physical control of the facility.

HOSTILE ACTION resulting in loss of physical control of the facility.

1) List site security shift supervision as Security Force.2) Added descriptors to better explain each safety function and allow for a timelyOperating Mode Applicability:

Operating Mode Applicability:

classification.

All 1,2,3,4, 5, DExample Emergency Action Levels: Emergency Action Level (EAL):1. a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift supervision).

1. A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.AND AND2. a. ANY Table HI safety function cannot be controlled or maintained.

b. EITHER of the following:

OR1. ANY of the following safety functions cannot be controlled or b. Damage to spent fuel has occurred or is IMMINENTmaintained.

• Reactivity control Table HI Safety Functions

• Core cooling [PIAR] / RPV water level [BRc CRCS heat removal -Reactivity Control(ability to shutdown the reactor and keep it shutdown)

OR -RPV Water Level (ability to cool the core)2. Damage to spent fuel has occurred or is IMMINENT

• RCS Heat Removal (ability to maintain a heatsink)

Page 52 of 66 EAL Comoarison Matrix DocumentNEI 99-M1 Rev 6 Proposed EAL Justification Initiating Condition:

SITE AREA EMERGENCY HOSTILE ACTION within the Protected Area.Operating Mode Applicability:

AllExample Emergency Action Levels:HSiHSiInitiating Condition:

HOSTILE ACTION within the Protected Area.Operating Mode Applicability:

No Change Difference Deviation

1) List site security shift supervision as Security Force.1,2,3,4,5, DEmergency Action Level (EAL):A HOSITLE ACTION is occurring or has occurred within the PROTECTED AREAas reported by the (site-security shift supervision).

A notification from the Security Force that a HOSTILE ACTION is occurring or hasoccurred within the PROTECTED AREA.Page 53 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL [ Justification HAIInitiating Condition:

ALERTHOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attackthreat within 30 minutes.Operating Mode Applicability:

AllExample Emergency Action Levels: (1 or 2)1. A HOSTILE ACTION is occurring or has occurred within the OWNERCONTROLLED AREA as reported by the (site-specific security shiftsupervision).

2. A validated notification from NRC of an aircraft attack threat within 30minutes of the site.HA1Initiating Condition:

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attackthreat within 30 minutes.Operating Mode Applicability:

1, 2, 3, 4, 5, DEmergency Action Level (EAL):1. A validated notification from NRC of an aircraft attack threat < 30 minutesfrom the site.OR2. Notification by the Security Force that a HOSTILE ACTION is occurring orhas occurred within the OWNER CONTROLED AREA.No Change 1 Difference 1-- Deviation

1) List site security shift supervision as Security Force.Page 54 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification HUI HUIDInitiating Condition:

UNUSUAL EVENT Initiating Condition:

H No Change M Difference FIDeviation Confirmed SECURITY CONDITION or threat.Confirmed SECURITY CONDITION or threat. 1) List site security shift supervision as Security Force.Operating Mode Appliicabiicabiity:

2) Further described credible security threat through listing a site-specific Operating Mprocedure.

All1,2,3,4,5, DExample Emergency Action Levels: (1 or 2 or 3)Emergency Action Level (EAL):1. A SECURITY CONDITION that does not involve a HOSTILE ACTION asreported by the (site-specific security shift supervision).

1. Notification of a credible security threat directed at the site as determined perSY-AA-101-132, Security Assessment and Response to Unusual Activities.

2. Notification of a credible security threat directed at the site. OR2. A validated notification from the NRC providing information of an aircraft3. A validated notification from the NRC providing information of an aircraft threat.threat.OR3. Notification by the Security Force of a SECURITY CONDITION that does notinvolve a HOSTILE ACTION.Page 55 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 1 Proposed EAL I Justification Initiating Condition:

SITE AREA EMERGENCY HS6Inability to control a key safety function from outside the Control Room.Operating Mode Applicability:

AllExample Emergency Action Levels: (1 and 2)Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that (site-specific number of minutes) hasbeen exceeded, or will likely be exceeded.

1. An event has resulted in plant control being transferred from the ControlRoom to (site-specific remote shutdown panels and local control stations).

2. Control of ANY of the following key safety functions is not reestablished within (site-specific number of minutes).

• Reactivity control* Core cooling [PWR] / RPV water level [BWR]RCS heat removalHS2Initiating Condition:

Inability to control a key safety function from outside the Control Room.Operating Mode Applicability:

1, 2, 3, 4, 5, DEmergency Action Level (EAL):Note:* The Emergency Director should declare the event promptly upon determining thatthe applicable time has been exceeded, or will likely be exceeded.

1 A Control Room evacuation has resulted in plant control being transferred from theControl Room to alternate locations per" SE-1, Plant Shutdown from the Remote Shutdown Panel -Procedure OR" SE-10, Plant Shutdown from the Alternative Shutdown Panels -Procedure AND2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.FD1 No Change M Difference I Deviation

1) EAL uses the site-specific Control Room evacuation procedures toeffectively list all of the alternate locations, panels, and stations requested bythe developer notes. This would be the procedures the Control Room wouldenter should such an event occur, this allows for greater clarity as to when theEAL would apply than if each panel and station used in alternate shutdownwere to be listed.2) Added descriptors to better explain each safety function and allow for atimely classification.

3) Changed "An event" to" A Control Room evacuation" to remove confusion ifpartial plant control was transferred to outside the Control Room with theControl Room still manned, due to testing or equipment failure.Table H1 Safety Functions

" Reactivity Control(ability to shutdown the reactor and keep it shutdown)

" RPV Water Level (ability to cool the core)" RCS Heat Removal (ability to maintain a heatsink)

Page 56 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification HAS HA2 m I1 I-Initiating Condition:

ALERT Initiating Condition:

LJ No Change Difference Deviation Control Room evacuation resulting in transfer of plant control to alternate locations.

Control Room evacuation resulting in transfer of plant control to alternate locations.

1) EAL uses the site-specific Control Room evacuation procedures to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedures the Control Room would enter should such anOperating Mode Applicability:

Operating Mode Applicability:

event occur, this allows for greater clarity as to when this EAL would apply than ifeach panel and station used in alternate shutdown were to be listed.All 1, 2, 3, 4, 5, D 2) Changed "An event" to" A Control Room evacuation" to remove confusion ifpartial plant control was transferred to outside the Control Room with the ControlExample Emergency Action Levels: Emergency Action Level (EAL): Room still manned, due to testing or equipment failure.An event has resulted in plant control being transferred from the Control Room to A Control Room evacuation has resulted in plant control being transferred from(site-specific remote shutdown panels and local control stations).

the Control Room to alternate locations per:" SE-1, Plant Shutdown from the Remote Shutdown Panel -Procedure OR" SE-1 0, Plant Shutdown from the Alternative Shutdown Panels -Procedure Page 57 of 66 EAL Comparison Matrix DocumentNE199-01 Rev 6 Proposed EAL Justification HU4HU3Initiating Condition:

UNUSUAL EVENTFIRE potentially degrading the level of safety of the plant.Operating Mode Applicability:

Initiating Condition:

FIRE potentially degrading the level of safety of the plant.Operating Mode Applicability:

1, 2, 3, 4, 5, DM No Change I7J Difference

--Deviation

1) Listed site-specific list of plant rooms or areas that contain SAFETYSYSTEM equipment to ensure timely classification.

AllExample Emergency Action Levels: (1 or 2 or 3 or 4)Note: The Emergency Director should declare the Unusual Eventpromptly upon determining that the applicable time has beenexceeded, or will likely be exceeded.

a. A FIRE is NOT extinguished within 15-minutes of ANY of thefollowing FIRE detection indications:

• Report from the field (i.e., visual observation)

• Receipt of multiple (more than 1) fire alarms or indications

• Field verification of a single fire alarmANDb. The FIRE is located within ANY of the following plant rooms orareas:(site-specific list of plant rooms or areas)2. a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).ANDb. The FIRE is located within ANY of the following plant rooms orareas:(site-specific list of plant rooms or areas)ANDc. The existence of a FIRE is not verified within 30-minutes of alarmreceipt.3. A FIRE within the plant or ISFSI [for plants with an ISFSI outside theplant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.

4. A FIRE within the plant or ISFSI [for plants with an ISFSI outside theplant Protected Area] PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

Emergency Action Level (EAL):Note:The Emergency Director should declare the event promptly upon determining that theapplicable time has been exceeded, or will likely be exceeded.

1. A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of thefollowing FIRE detection indications:

• Report from the field (i.e., visual observation)

• Receipt of multiple (more than 1) fire alarms or indications

• Field verification of a single fire alarmTable H2 Vital Areas* Reactor Building (when inerted the Drywell is exempt)* Control Room* Diesel Generator Building* Emergency Pump Structure

• Inner Screen Structure

• Emergency Cooling Tower* Emergency Switchgear/Battery Rooms* Cable Spread RoomOR2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of aFIRE).ANDb. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.OR3. A FIRE within the plant PROTECTED AREA or ISFSI PROTECTED AREA notextinguished in < 60-minutes of the initial report, alarm or indication.

OR4. A FIRE within the plant PROTECTED AREA or ISFSI PROTECTED AREA thatrequires firefighting support by an offsite fire response agency to extinguish.

Page 58 of 66 EAL Cornoarison Matrix DocumentNEI 99-01 Rev 6Proposed EALJustification Initiating Condition:

UNUSUAL EVENTSeismic event greater than OBE levels.Operating Mode Applicability:

HU2HU4Initiating Condition:

Seismic event greater than OBE levels.Operating Mode Applicability:

1, 2, 3,4, 5, DM No Change 1 Difference Deviation

1) Listed site-specific indications.

AllExample Emergency Action Levels:Seismic event greater than Operating Basis Earthquake (OBE) as indicated by:a. (site-specific indication that a seismic event met or exceeded OBE limits)Emergency Action Level (EAL):Seismic event > Operating Basis Earthquake (OBE) as indicated by the Alarmon Panel 316 A-5,"OPERATING BASIS EARTHQUAKE EXCEEDED" Page 59 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification HA5 HA5mmmInitiating Condition:

ALERT Initiating Condition:

[AS No Change LJ Difference L Deviation Gaseous release impeding access to equipment necessary for normal plant Gaseous release impeding access to equipment necessary for normal plant operations,

1) Listed plant specific rooms and areas with entry related mode applicability tooperations, cooldown or shutdown, cooldown or shutdown, ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

All 1, 2,3, 4,5, DExample Emergency Action Levels: Emergency Action Level (EAL):Note: If the equipment in the listed room or area was already inoperable, or out Note:of service, before the event occurred, then no emergency classification iswarranted.

• If the equipment in the listed room or area was already inoperable, or out ofservice, before the event occurred, then no emergency classification is1. a. Release of a toxic, corrosive, asphyxiant or flammable gas into any warranted.

of the following plant rooms or areas:(site-specific list of plant rooms or areas with entry-related mode 1. Release of a toxic, corrosive, asphyxiant or flammable gas in a Table H3 area.applicability identified)

Table H3AND Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability

b. Entry into the room or area is prohibited or impeded.Control Room Modes 1, 2, 3, 4, and 5Reactor Building*

Modes 3, 4 and 5-Areas required to establish shutdown coolingAND2. Entry into the room or area is prohibited or impeded.Page 60 of 66 FAL Comnaricon Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENTHazardous EventOperating Mode Applicability:

HU3AllExample Emergency Action Levels: (1 or 2 or 3 or 4)Note: EAL #3 does not apply to routine traffic impediments such as fog, snow,ice, or vehicle breakdowns or accidents.

I. A tornado strike within the PROTECTED AREA.2. Internal room or area flooding of a magnitude sufficient to require manualor automatic electrical isolation of a SAFETY SYSTEM component neededfor the current operating mode.3. Movement of personnel within the PROTECTED AREA is impeded due toan offsite event involving hazardous materials (e.g., an offsite chemicalspill or toxic gas release).

4. A hazardous event that results in on-site conditions sufficient to prohibitthe plant staff from accessing the site via personal vehicles.

5. (Site-specific list of natural or technological hazard events)HU6Initiating Condition:

Hazardous EventOperating Mode Applicability:

1, 2, 3, 4, 5, DEmergency Action Level (EAL):Note:* EAL #4 does not apply to routine traffic impediments such as fog, snow,ice, or vehicle breakdowns or accidents.

1. Tornado strike within the PROTECTED AREA.OR2. Internal room or area flooding of a magnitude sufficient to require manual orautomatic electrical isolation of a SAFETY SYSTEM component required byTechnical Specifications for the current operating mode.OR3. Movement of personnel within the PROTECTED AREA is impeded due to anoffsite event involving hazardous materials (e.g., an offsite chemical spill ortoxic gas release).

OR4. A hazardous event that results in on-site conditions sufficient to prohibit theplant staff from accessing the site via personal vehicles.

OR5. Abnormal River level, as indicated by EITHER:* > 112 ft. (high level)OR* < 98.5 ft. (low level)D] No Change FIDifference F Deviation

1) Included river water level as part of the site-specific list of natural ortechnological hazard events. The EAL values selected are the current ApprovedUE EAL values.2) Changed the word "needed" to "required by Technical Specifications" in the EALto be consistent with terminology used by operators and minimize confusion.

Page 61 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 ] Proposed EAL I Justification HG7HG7Initiating Condition:

GENERAL EMERGENCY Initiating Condition:

M No Change 1 Difference FIDeviation Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a General Emergency.

Operating Mode Applicability:

AllExample Emergency Action Levels:Other conditions exist which in the judgment of the Emergency Director indicatethat events are in progress or have occurred which involve actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control ofthe facility.

Releases can be reasonably expected to exceed EPA Protective ActionGuideline exposure levels off-site for more than the immediate site area.Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a General Emergency.

Operating Mode Applicability:

1,2, 3,4, 5, DEmergency Action Level (EAL):Other conditions exist which in the judgment of the Emergency Director indicatethat events are in progress or have occurred which involve actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control ofthe facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area..Page 62 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification HS7 HS7 FInitiating Condition:

SITE AREA EMERGENCY Initiating Condition:

LX--1 No Change j Difference

-Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of a Site Area Emergency.

declaration of a Site Area Emergency.

Operating Mode Applicability:

Operating Mode Applicability:

All 1, 2, 3,4, 5, DExample Emergency Action Levels: Emergency Action Level (EAL):Other conditions exist which in the judgment of the Emergency Director indicate Other conditions exist which in the judgment of the Emergency Director indicatethat events are in progress or have occurred which involve actual or likely major thaents e in the jud which Emerge a ctor indicatefailures of plant functions needed for protection of the public or HOSTILE ACTION ai events are in progress or have occurred which involve actual or likely majorthat results in intentional damage or malicious acts; (1) toward site personnel or tures of plant functions needed for protection of the public or HOSTILE ACTIONequipment that could lead to the likely failure of or; (2) that prevent effective that results in intentional damage or malicious acts; (1) toward site personnel oraccess to equipment needed for the protection of the public. Any releases are not equipment that could lead to the likely failure of or; (2) that prevent effective expected to result in exposure levels which exceed EPA Protective Action access to equipment needed for the protection of the public. Any releases are notGuideline exposure levels beyond the site boundary o expected to result in exposure levels which exceed EPA Protective ActionGuideline exposure levels beyond the site boundary.

Page 63 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification HAS HA7 NInitiating Condition:

ALERT Initiating Condition:

No Change Difference Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrantdeclaration of an Alert. declaration of an Alert.Operating Mode Applicability:

Operating Mode Applicability:

All 1,2, 3,4, 5, DExample Emergency Action Levels: Emergency Action Level (EAL):Other conditions exist which in the judgment of the Emergency Director indicate Other conditions exist which in the judgment of the Emergency Director indicatethat events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that that events are in progress or have occurred which involve an actual or potential involves probable life threatening risk to site personnel or damage to site substantial degradation of the level of safety of the plant or a security event thatequipment because of HOSTILE ACTION. Any releases are expected to be involves probable life threatening risk to site personnel or damage to sitelimited to small fractions of the EPA Protective Action Guideline exposure levels, equipment because of HOSTILE ACTION. Any releases are expected to belimited to small fractions of the EPA Protective Action Guideline exposure levels.Page 64 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification l-U7 Hu7 H --Initiating Condition:

UNUSUAL EVENT Initiating Condition:

F'7 No Change Difference Deviation Other conditions existing which in the judgment of the Emergency director warrantdeclaration of an UNUSUAL EVENT. Other conditions existing which in the judgment of the Emergency director warrantdeclaration of an UNUSUAL EVENT.Operating Mode Applicability:

Operating Mode Applicability:

All1,2,3,4.5, DExample Emergency Action Levels:Other conditions exist which in the judgment of the Emergency Director indicate Emergency Action Level (EAL):that events are in progress or have occurred which indicate a potential Other conditions exist which in the judgment of the Emergency Director indicatedegradation of the level of safety of the plant or indicate a security threat to facility that events are in progress or have occurred which indicate a potential protection has been initiated.

No releases of radioactive material requiring offsite degradation of the level of safety of the plant or indicate a security threat to facilityresponse or monitoring are expected unless further degradation of safety systems protection has been initiated.

No releases of radioactive material requiring offsiteoccurs. response or monitoring are expected unless further degradation of safety systemsoccurs.Page 65 of 66 EAL Comparison Matrix DocumentNEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENTDamage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability:

E-HUIAllExample Emergency Action Levels:Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specifictechnical specification allowable radiation level) on the surface of the spent fuelcask.E-HUIInitiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability:

1,2,3,4,5, DEmergency Action Level (EAL):Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:* > 350 mr/hr gamma and 30 mr/hr neutron on the top (protective cover).OR* > 250 mrlhr gamma and 60 mrlhr neutron on the sides of the radialneutron shield.OR* > 1100 mrlhr gamma and 250 mr/hr neutron on the side surfaces of thecask above the redial neutron shield.OR* > 800 mr/hr gamma and 600 mr/hr neutron on the side surfaces of thecask below the radial neutron shield.I-H No Change FIDifference FIDeviation

1) Listed 2x the site-specific cask specific allowable radiation level as perCertificate of Compliance No. 1014 Appendix A Section 5.7.Page 66 of 66 NuclearPe~ach Rnttnm Atomic Power Stsation Annex Exelon NucilearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARGIInitiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRfem TEDEor 5000 mRr-em thyroid CDE.Operating Mode Applicability:

1, 2, 3,4, 5, DEmergency Action Level (EAL):Notes:" The Emergency Director should declare the General .- event promptly upondetermining that the applicable time has been exceeded, or will likely be exceeded.

" If an ongoing release is detected and the release start time is unknown, assume thatthe release duration has exceeded 15 minutes." Classification based on effluent monitor readings assumes that a release path to theenvironment is established.

If the effluent flow past an effluent monitor is known tohave stopped due to actions to isolate the release path, then the effluent monitorreading is no longer valid for classification purposes.

" The pre-calculated effluent monitor values presented in EAL #1 should be used foremergency classification assessments until the results from a dose assessment usingactual meteorology are available.

(1) Reading on 'A *NY of t~he following radiation monitors greater than the reading shownfr15- minute or logr(site monitor list and threshold 1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 minutes:[ Table RI Effluent Monitor Thresholds Release Path General Emergency Main Stack (RI-0-17-050B Common) 5.57 E+09 uCi/secVent Stack (RI-2979B Unit 2 or RI- 1.60 E+08 uCi/sec3979B Unit 3)OR2. Dose assessment using actual meteorology indicates doses at or beyond (rate speriflrdose receptor point) the site boundary of EITHER:a. > 1000 mRem TEDEMonth 20XXPBAPS 3-1EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ORb. > 5000 mRem CDE ThyroidORP IJi-lola EUP.~'O'2' resuirn inoicaro

~m m in-i~i*c at me moilowino at or DO~Ofl0 simo ceocilic--------------------------------

~5~Gdo06 point).Cl-o,,ed, window dose rates greater than 1,000 mR'hr expected to continuefor 60 minuts Or longer.Analyses nof f4ield rurvey samples indicate thyroid CDE greater than 5,000 mr~emfor ono hour ofinlaon3. Field survey results at or beyond the site boundary indicate EITHER:a. Gamma (closed window) dose rates >1000 mR/hr are expected to continuefor > 60 minutes.ORb. Analyses of field survey samples indicate

> 5000 mRem CDE Thyroid for60 minutes of inhalation.

Basis:This IC addresses a release of gaseous radioactivity that results in projected or actualoffsite doses greater than or equal to the EPA Protective Action Guides (PAGs). Itincludes both monitored and un-monitored releases.

Releases of this magnitude willrequire implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plantconditions alone. The inclusion of both plant condition and radiological effluent EALsmore fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1000 mRfem while the 5000 mRrem thyroidCDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE andthyroid CDE.Classific~ation based on effluent moenitor readings assumes that a release path to thenViFronment is established.

if the effluent flow past an effluent moenitor is, known to havstopped due to actions to iso0latte the release path, then the efflucnt moniorF reading is nolGRner valid for classification nrWo~ses.

Basis Reference(s):

Month 20XXPBAPS 3-2EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

1. NEI 99-01 Rev 6, AG12. Peach Bottom ODCM3. EP-EAL-0607 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EALThreshold Values Peach Bottom Atomic Power Station4. EP-AA-112-500, Emergency Environmental Monitoring Month 20XXPBAPS 3-3EP-AA-1007 (Revision XX)

Pone-h P^++^rn Afernid-Pnuuor Qfafinn AnnoyIPvlnn Nii-lo,,r D~~rk Rn44r~m A$r~mii.

Br~ai~r ~$~*rnn A nn~v Fv~Inn kiuiria2r RECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARSlInitiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mRfem TEDEor 500 mRfem thyroid CDE.Operating Mode Applicability:

1, 2, 3, 4, 5, DEmergency Action Level (EAL):Notes:" The Emergency Director should declare the Site Area E=m.ergency event promptlyupon determining that the applicable time has been exceeded, or will likely beexceeded.

• If an ongoing release is detected and the release start time is unknown, assume thatthe release duration has exceeded 15 minutes." Classification based on effluent monitor readings assumes that a release path to theenvironment is established.

If the effluent flow past an effluent monitor is known tohave stopped due to actions to isolate the release path, then the effluent monitorreading is no longer valid for classification purposes.

" The pre-calculated effluent monitor values presented in EAL #1 should be used foremergency classification assessments until the results from a dose assessment usingactual meteorology are available.

(1) Reading on ANY of the radiation m.ontors greater than the reading shownfo-r 159 minu6te rm lon!ger:9 tb!M butibamatj HWWWRUM ambt d!!U tHitibHUOU vdouutrt(2) Dose assessment using actual mneteorology indicates doses greater than 10-0Mmre TE-DE o-r 150 mrem thyroid CDPE at orF beyond (Site specific.

dose recepto(3) Field sur'ey results indicate EITHEPR of the fallow.ing at or beyond (site specificdoe r .cepto point):" Closed window dose r-ates greAter-than 100 miRih expected to ccntinue for- 60 minutesor efioger-" Analyses of field survey samples indicate thyroid CDE greater-than 500 wremn for- oefhour- of inhalation.

Month 20XXPBAPS 3-4EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 minutes:Table R1 Effluent Monitor Thresholds Release Path Site Area Emergency Main Stack (RI-0-17-050B Common) 5.57 E+08 uCi/secVent Stack (RI-297/B Unit 2 or RI- 1.60 E+07 uCi/sec3979B Unit 3)OR2. Dose assessment using actual meteorology indicates doses at or beyond the siteboundary of EITHER:a. > 100 mRem TEDEORb. > 500 mRem CDE ThyroidOR3. Field survey results at or beyond the site boundary indicate EITHER:a. Gamma (closed window) dose rates >100 mR/hr are expected to continuefor > 60 minutes.ORb. Analyses of field survey samples indicate

> 500 mRem CDE Thyroid for 60minutes of inhalation.

Basis:This IC addresses a release of gaseous radioactivity that results in projected or actualoffsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs).It includes both monitored and un-monitored releases.

Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plantconditions alone. The inclusion of both plant condition and radiological effluent EALsmore fully addresses the spectrum of possible accident events and conditions.

Month 20XXPBAPS 3-5EP-AA-1007 (Revision XX)

Exelon NuclearPesach Rnttnm Atomic~ Power Staition Annex Exelnn Nuclea~rRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS The TEDE dose is set at 10% of the EPA PAG of 1000 mRrem while the 500 mRremthyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDEand thyroid CDE.Classific-ation bhased- on effluWent Monitor readin~gs assumes that a release path tothe envronmRFent is established.

if the effluent flow past an effluent monitor isknown tohave stopped due to- ac-tion to 1 the release path, then the monior redingir, no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RAG1.Basis Reference(s):

1. NEI 99-01 Rev 6, AS12. Peach Bottom ODCM3. EP-EAL-0607 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EALThreshold Values Peach Bottom Atomic Power Station4. EP-AA-1 12-500, Emergency Environmental Monitoring Month 20XXPBAPS 3-6EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA1Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mR-remTEDE or 50 mRfem thyroid CDE.Operating Mode Applicability:

1, 2,3, 4,5, DEmergency Action Level (EAL):Notes:* The Emergency Director should declare the AleFt-event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

• If an ongoing release is detected and the release start time is unknown, assume thatthe release duration has exceeded 15 minutes.* Classification based on effluent monitor readings assumes that a release path to theenvironment is established.

If the effluent flow past an effluent monitor is known tohave stopped due to actions to isolate the release path, then the effluent monitorreading is no longer valid for classification purposes.

9 The pre-calculated effluent monitor values presented in EAL #1 should be used foremergency classification assessments until the results from a dose assessment usingactual meteorology are available.

/4\ D ,,; .; n AKlV f'k fAh A,-;ll,-+

kk,;n, r.,A;,t;,n r.,-n;÷r.+

,ra-,r ÷k ,n ÷a ra+inn ohnt.,+ AforF 15 minutes or longer:(site specific monmitor list and threshold values)(2) Dose assessment using ac~tuaal meateorology idctsdoese greater thnn 10Q mremTE=DE Or 50 mre8Rn thyroid CDE at or beyon~d (Site GpeciG& dose receptor point)-.(3) Analysis of a liquid effluent sample indicates a concentration or release rate thatwould- resul-t in doses greater than 10Q mArem TEDE or 50 mrem thyroid-CDE atobeyond (site specific dose receptor point) forF one hour Of eXPoSUre(4) Field survey Fesults, indicsate EITHE~R_

of the following at or beyond (site specificdose receptor point):;" Closed Aindow dose rates gretere than! 10 m+R'h expeeted to continue for 60 mninutes or-" Analyses of field sur1'ey samples indicate thyroeid CDE gr-eater than 50 mfeem for- on-ehour of inhalation.

Month 20XXPBAPS 3-7EP-AA-1007 (Revision XX)

NuclearPevwach vnttmm Atonmicm Power Stvation Annex Ex.lon NucleamrRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS

1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 minutes.Table R1 Effluent Monitor Thresholds Release Path AlertMain Stack (RI-0-17-050B Common) 5.57 E+07 uCi/secVent Stack (RI-2979B Unit 2 or RI- 1.60 E+06 uCi/sec3979B Unit 3)OR2. Dose assessment using actual meteorology indicates doses at or beyond the siteboundary of EITHER:a. > 10 mRem TEDEORb. > 50 mRem CDE ThyroidOR3. Analysis of a liquid effluent sample indicates a concentration or release rate thatwould result in doses greater than EITHER of the following at or beyond the siteboundary:

a. 10 mRem TEDE for 60 minutes of exposureORb. 50 mRem CDE Thyroid for 60 minutes of exposureOR4. Field survey results at or beyond the site boundary indicate EITHER:a. Gamma (closed window) dose rates > 10 mR/hr are expected tocontinue for a 60 minutes.ORMonth 20XXPBAPS 3-8EP-AA-1007 (Revision XX)

Peach Boftom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

b. Analyses of field survey samples indicate

> 50 mRem CDEThyroid for 60 minutes of inhalation.

Basis:This IC addresses a release of gaseous or liquid radioactivity that results in projected oractual offsite doses greater than or equal to 1% of the EPA Protective Action Guides(PAGs). It includes both monitored and un-monitored releases.

Releases of thismagnitude represent an actual or potential substantial degradation of the level of safety ofthe plant as indicated by a radiological release that significantly exceeds regulatory limits(e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plantconditions alone. The inclusion of both plant condition and radiological effluent EALsmore fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1000 mRfem while the 50 mR-em thyroidCDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE andthyroid CDE.Classificatfion based on effluent mRonitor readings assumes that a release path tothe enViFronment is established.

if the efflun flo pst an offlUcnAt monRitorF is knownthave stopped due to actions, to- isolat-e the release path, then the effluent moniRtor readingisR no lonoer vali forg c~assificationnn

-Duroer,.

Escalation of the emergency classification level would be via IC RAS1.Basis Reference(s):

1. NEI 99-01 Rev 6, AA12. Peach Bottom ODCM3. EP-EAL-0607 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EALThreshold Values Peach Bottom Atomic Power Station4. EP-EAL-0614Revision 0 Peach Bottom Criteria for Choosing Radiological LiquidEffluent EAL Threshold ValuesMonth 20XXPBAPS 3-9EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARUMInitiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM (Site speGifteffluent release controlling document) limits for 60 minutes or longer.Operating Mode Applicability:

1, 2, 3, 4, 5, DEmergency Action Level (EAL):Notes:" The Emergency Director should declare the Unusual Event cvent promptly upondetermining that 60 minutesthe applicable time has been exceeded, or will likely beexceeded.

" If an ongoing release is detected and the release start time is unknown, assume thatthe release duration has exceeded 60 minutes." Classification based on effluent monitor readings assumes that a release path to theenvironment is established.

If the effluent flow past an effluent monitor is known tohave stopped due to actions to isolate the release path, then the effluent monitorreading is no longer valid for classification purposes.

(1) Reading on ANY effluent radiation monitor greater than 2 times the (site specifieffluent Felease conRtrolling document) limits for 60 minutes Or !ogcR:e(site specific mroitor liSt and threshold ValUes cor.esponding to 2 times thecontro÷9lliRg lImitf)(2) Reading On A NY e-ffluenA-t radiation moni~tor greater than 2 times the alarm 6etpinestabIse b IIy a current radioactivity discharge permit for 60 min;utes Or longer.(3) Sample analysis for a gaseou-is or liquid reloaso-indic-atees a concentration orreleAaseA rate greater than :2 times the (site specific efflu ent release controlling document) imits forF 60 minutes OrF Ilger.1. Reading on any of the following effluent monitors

> 2 times alarm setpointestablished by a current radioactive release discharge permit for > 60 minutes." Radwaste Discharge Effluent Monitor (RR63-OR001)

OR" Discharge Permit specified monitorORMonth 20XXPBAPS 3-10EP-AA-1007 (Revision XX)

Pant-h nffnm Afnmie Pnwar lUmfinn AnnovIP::letIn km,,l,-or Da2vh RM*r~m A*nmi, Dr~wcr ~f~*ir~n A nn~v Fvalnn M..r~Inar RECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. Readings on ANY Table R1 Effluent Monitor > Table RI value for > 60 minutes:Table R1 Effluent Monitor Thresholds Release Path I Unusual EventMain Stack 6.36 E+05 uCi/sec(RI-0-17-050B Common)Main Stack 6.36 E+05 uCi/sec(RI-0-17-050A Common)Vent Stack (RI-2979B Unit 2 orRI-3979B Unit 3) 1.82 E+04 uCi/secVent Stack (RI-2979A Unit 2 orRI-3979A Unit 3) 1.82 E+04 uCi/secOR3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.Basis:This IC addresses a potential decrease in the level of safety of the plant as indicated by alow-level radiological release that exceeds regulatory commitments for an extendedperiod of time (e.g., an uncontrolled release).

It includes any gaseous or liquidradiological

release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release ofradioactive effluents to the environment.

Further, there are administrative controlsestablished to prevent unintentional

releases, and to control and monitor intentional releases.

The occurrence of an extended, uncontrolled radioactive release to theenvironment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events andconditions that cannot be readily or appropriately classified on the basis of plantconditions alone. The inclusion of both plant condition and radiological effluent EALsmore fully addresses the spectrum of possible accident events and conditions.

CQassifiatien based on effluent monitor readings assumes that a release pathtthe "A-omt ieb dif the efflucnt flow past -an effluent monito is known thave stopped due to acrtions to lt the release path, then the effluent monito-r r-ead.i-ng s n, oR longer valid foFr lassifi-ati;n rposes.Releases should not be prorated or averaged.

For example, a release exceeding 4 timesrelease limits for 30 minutes does not meet the EAL.EAL #1 Basis: IMonth 20XXPBAPS 3-11EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexI::alnn N",AlnnrRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS FAL--##--

This EAL addresses radioactivity releases that cause effluent radiation monitorreadings to exceed 2 times the limit established by a radioactivity discharge permit. ThisEAL will typically be associated with planned batch releases from non-continuous releasepathways (e.g., radwaste, waste gas).The effluent monitors listed are those normally used for planned discharges.

If adischarge is performed using a different flowpath or effluent monitor other than thoselisted (e.g., a portable or temporary effluent monitor),

then the declaration criteria will bebased on the monitor specified in the Discharge Permit.EAL #2 Basis:EAL--#-1-This EAL addresses normally occurring continuous radioactivity releases frommonitored gaseous .,liqud- effluent pathways.

EAL #3 Basis:FAL- #3 -This EAL addresses uncontrolled gaseous or liquid releases that are detectedby sample analyses or environmental

surveys, particularly on unmonitored pathways(e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river watersystems, etc.).Escalation of the emergency classification level would be via IC RAA1.Basis Reference(s):

1. NEI 99-01 Rev 6, AU12. Peach Bottom ODCM3. EP-EAL-0607 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EALThreshold Values Peach Bottom Atomic Power StationMonth 20XXPBAPS 3-12EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA2 IInitiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability:

1, 2,3,4,5, DEmergency Action Level (EAL):(1) IUncove; of irradiated fuel in t-heD REFUEI1 ING PADTHWVAY.V (2) Damnage to irradiated fue! resulting in a release of radioactivity fromn the fuel asInd;.ated by ANY of the radiation onRit..r:

(c~ite specific listing of Fadiation

monitors, and the associated

readings, sctpoints and/or alars)(3) Lowering of spent fuel pool level to (site specifc Level 2 value). [See DevoIlp1. Uncovery of irradiated fuel in the REFUELING PATHWAY.OR2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel asindicated by ANY Table R2 Radiation Monitor reading >1000 mRemlhrTable R2Refuel Floor ARM's* Steam Separator Pool, 234 ELEV, 3.7 (7.9)" Refuel Slot, 234 ELEV, 3.8 (7.10)" Fuel Pool, 234 ELEV, 3.9 (7.11)* Refueling Bridge, 234 ELEV, 3.10 (7.12)Month 20XXPBAPS 3-13EP-AA-1007 (Revision XX)

Pont-h nftnm Afnrnit-Pnwor Atnfinn AnnoyFvolnn N~nelonrPa~2v~II RnI*nm Atnmi,~ Pnwgir ~t~a$inn Ann~v Fvi~Inn N.ur~Ic~r RECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Basis:REFUELING PATHWAY:

all the cavities, tubes, canals and pools through whichirradiated fuel may be moved or stored, but not including the reactor vessel below theflange.IMMINENT:

The trajectory of events or conditions is such that an EAL will be met within arelatively short period of time regardless of mitigation or corrective actions.CONFINEMENT BOUNDARY:

The irradiated fuel dry storage cask barrier(s) betweenareas containing radioactive substances and the environment.

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly.,

or a sinificant lo ...ing of watert ol w.it.hin th9 spnt f.4el p90l ...

Nots). These events present radiological safety challenges to plantpersonnel and are precursors to a release of radioactivity to the environment.

As such,they represent an actual or potential substantial degradation of the level of safety of theplant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that theloaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss ofthe CONFINEMENT BOUNDARY is classified in accordance with IC E-HUI.Escalation of the emergency would be based on either Recognition Category A or C ICs.EAL #1 Basis:E=AL-#-tThis EAL escalates from RAU2 in that the loss of level, in the affected portion of theREFUELING

PATHWAY, is of sufficient magnitude to have resulted in uncovery ofirradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visualobservation (e.g., reports from personnel or camera images),

as well as significant changes in water and radiation levels, or other plant parameters.

-If available, videocameras may allow remote observation.

Computational aids may also be used (e.g., aboil-off curve). Classification of an event using this EAL should be based on the totality ofavailable indications, reports and observations.

While an area radiation monitor could detect aa rise in a dose rate due to alowering of water level in some portion of the REFUELING

PATHWAY, the reading maynot be a reliable indication of whether or not the fuel is actually uncovered.

To the degreepossible, readings should be considered in combination with other available indications ofinventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified inaccordance Recognition Category C during the Cold Shutdown and Refueling modes.Month 20XXPBAPS 3-14EP-AA-1007 (Revision XX)

Ppnn~h Rnfttm Atomic~ Power Stsation Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS EAL #2 Basis:RAI #2This EAL addresses a release of radioactive material caused by mechanical damage toirradiated fuel. Damaging events may include the dropping, bumping or binding of anassembly, or dropping a heavy load onto an assembly.

A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of apotential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency would be based on either Recognition Category RA-or CICs.Spont fuel pool water level at this Yalue is within the lower end- of the love! rangenecessar,'

to prevent signfificant doso conrequenc-es from. direct gamma radiation to perfoer;ing operations, in the Viinity of the spent fuel pool. This condition

, a significant losr, of spent fuel pool waater inventor; and thus it i"s also a precursor to a loss of the ability to adequately cool the irradiated fuel assemnbles stored in; the polEscnalatfionl of the em~ergency c~lassAific-ation level wou-id b-e via lGS A8i or A82 (seeASDeveloe oo)IBasis Reference(s):

1. NEI 99-01 Rev 6, AA22. ON-124, Fuel Floor and Fuel Handling Problems3. P-S-43, Radiation Monitoring System4. ARC 003-00C214 B-4, Refueling Floor Area Hi Radiation

5. ARC 20C075 C-1, Fuel Storage Pool Hi Radiation Month 20XXPBAPS 3-15EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARU2Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability:

1, 2,3,4,5, DEmergency Action Level (EAL):(1) a. UNPI'INlINE-D tr I'tFlel drop in the RDEFUIELINGIhlQ-PATHIAIAV as indicated byANYof the fellow ng_(site specaific.

level indications).

b. UNPLANNED rie inae adiation levels as indicated by ANY of thefollowi~ng radiation moenitor-s.

(sito speci*fic-list of area radiation monitors) 1 .a .UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

" Refueling Cavity water level < 458 inches on Shutdown/Refuel Rangeindicator LI-2(3)-2-3-86 or PRlFR-2(3)-06-097.

OR* Spent Fuel Pool level < 232 feet 3 inches plant elevation.

OR1 Indication or report of a drop in water level in the REFUELING PATHWAY.ANDb. UNPLANNED Area Radiation Monitor reading rise on one or more radiation monitors in Table R2.Month 20XXPBAPS 3-16EP-AA-1007 (Revision XX)

Peach Boftom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Table R2Refuel Floor ARM's" Steam Separator Pool, 234 ELEV, 3.7 (7.9)" Refuel Slot, 234 ELEV, 3.8 (7.10)* Fuel Pool, 234 ELEV, 3.9 (7.11)" Refueling Bridge, 234 ELEV, 3.10 (7.12)Basis:UNPLANNED:

A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.REFUELING PATHWAY:

all the cavities, tubes, canals and pools through whichirradiated fuel may be moved or stored, but not including the reactor vessel below theflange.This IC addresses a deGiease-loss in water level above irradiated fuel sufficient to causeelevated radiation levels. This condition could be a precursor to a more serious eventand is also indicative of a minor loss in the ability to control radiation levels within theplant. It is therefore a potential degradation in the level of safety of the plant.A water level derease-loss will be primarily determined by indications from available levelinstrumentation.

Other sources of level indications may include reports from plantpersonnel (e.g., from a refueling crew) or video camera observations (if available) or fromany other temporarily installed monitoring instrumentation.

A significant drop in the waterlevel may also cause an irGeasea rise in the radiation levels of adjacent areas that canbe detected by monitors in those locations.

The effects of planned evolutions should be considered.

For example, a refueling bridgearea radiation monitor reading may-iPeaserise due to planned evolutions such as liftingof the reactor vessel head or movement of a fuel assembly.

Note that this EAL isapplicable only in cases where the elevated reading is due to an UNPLANNED loss ofwater level.A drop in water level above irradiated fuel within the reactor vessel may be classified inaccordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the emergency classification level would be via IC RAA2.Basis Reference(s):

1. NEI 99-01 Rev 6, AU2Month 20XXPBAPS 3-17EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. Technical Specifications 3.9.6, Reactor Pressure Vessel (RPV) Water Level3. ON-124, Fuel Floor and Fuel Handling Problems4. P-S-43, Radiation Monitoring System5. ARC 003-00C214 B-4, Refueling Floor Area Hi Radiation

6. ARC 20C075 C-1, Fuel Storage Pool Hi Radiation

7. GP-6, Refueling Operations Month 20XXPBAPS 3-18EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA3Initiating Condition:

Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

Operating Mode Applicability:

1, 2, 3, 4, 5, DEmergency Action Level (EAL):Note: If the equipment in the listed-room or area listed in Table R4 was alreadyinoperable, or out of service, before the event occurred, then no emergency classification is warranted (1) Dose rate greater than 15 mR'~hr in ANY of the following areas:Control 'oom-Central Alarm Station9 (other site speeifle afeas/fooms)

(2) An UNPLA.NNED event results in radiation levels that prohibit or impede access toan" of the following plant rooms or areas:(site specific list of plant rooms Or areas with entr; related mode appkiability

1. Dose rate gFeateF-than>

15 mRlhr in ANY of the following Table R3 areas:Table R3Areas Requiring Continuous Occupancy

" Main Control Room" Central Alarm Station -(by survey)ORMonth 20XXPBAPS 3-19EP-AA-1007 (Revision XX)

IExolon NueloarPeacrh Bonttom Atomic~ Power Station Annex Exelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS

2. UNPLANNED event results in radiation levels that prohibit or significantly impedeaccess to any of the following Table R4 plant rooms or areas:Table R4Areas with Entry Related Mode Applicability Area Entry Related ModeApplicability Reactor Building*

Modes 3, 4 and 5*Areas required to establish shutdown coolingBasis:UNPLANNED:

A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.This IC addresses elevated radiation levels in certain plant rooms/areas sufficient topreclude or impede personnel from performing actions necessary to transition the plantfrom normal plant operation to cooldown and shutdown as specified in normal plantproceduresmaintain normal plant operation, a normal plant coldown and&h'-tdowR.

As such, it represents an actual or potential substantial degradation of thelevel of safety of the plant. The Emergency Director should consider the cause of theincreased radiation levels and determine if another IC may be applicable.

Table R4 is a list of plant rooms or areas with entry-related mode applicability that containequipment which require a manual/local action necessary to transition the plant fromnormal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling),

where if this action is not completed the plantwould not be able to attain and maintain cold shutdown.

This Table does not include rooms or areas for which entry is required solely to performactions of an administrative or record keeping nature (e.g., normal rounds or routineinspections).

Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including theControl Room.For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, ormay be, procedurally required during the plant operating mode in effect and the elevatedradiation levels preclude the ability to place shutdown cooling in serviceat the time of theelevated radiation levels. The emergency classification is not contingent upon whetherentry is actually necessary at the time of the increased radiation levels. Access should beconsidered as impeded if extraordinary measures are necessary to facilitate entry ofpersonnel into the affected room/area (e.g., installing temporary shielding beyond thatrequired by procedures, requiring use of non-routine protective equipment, requesting anMonth 20XXPBAPS 3-20EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS extension in dose limits beyond normal administrative limits).An emergency declaration is not warranted if any of the following conditions apply." The plant is in an operating mode different than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time ofthe elevated radiation levels).

For example, the plant is in Mode 1 when the radiation a Areaserise occurs, and the procedures used for normal operation, cooldown andshutdown do not require entry into the affected room until Mode 4." The increased radiation levels are a result of a planned activity that includescompensatory measures which address the temporary inaccessibility of a room orarea (e.g., radiography, spent filter or resin transfer, etc.)." The action for which room/area entry is required is of an administrative or recordkeeping nature (e.g., normal rounds or routine inspections).

" The access control measures are of a conservative or precautionary nature, andwould not actually prevent or impede a required action.Escalation of the emergency classification level would be via Recognition Category RA, Cor F ICs.Basis Reference(s):

1. NEI 99-01 Rev 6, AA32. FSAR Section describing Class I structures

3. Appendix R Analysis defining Safe Shutdown Area4. P-S-43, Radiation Monitoring System5. T-300, Fire GuidesMonth 20XXPBAPS 3-21EP-AA-1007 (Revision XX)

Pant-h ^+frmAfrmie-Pnuuar-QfnfienAnnov P::vainn k,,Pelaar Ba~t-h Rw~*fr~m A *nm iv~ Dnui~r ~*~*inn An n~v ~vc~Inn Nm.rIa~rRECOGNITION CATEGORYABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS SRU3Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):(1) (Site specific.

radiation monitor) reading greater than (site Specific value)-.(2) Sample analysis indicates that a reactor coolant activ.it' value is greater than anallowable limit specified in Technical Specifications.

1. Air Ejector discharge radiation monitor RIS-2(3)-17-150 A/B Hi-Hi alarm.OR2. Specific coolant activity

> 4.0 uCllgm Dose equivalent 1-131.Basis:This IC addresses a reactor coolant activity value that exceeds an allowable limitspecified in Technical Specifications.

This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.Conditions that cause the specified monitor to alarm that are not related to fuel claddegradation should not result in the declaration of an Unusual Event.This EAL addresses site-specific radiation monitor readings that provide indication of adegradation of fuel clad integrity.

An Unusual Event is only warranted when actual fuel clad damage is the cause of theelevated coolant sample activity (as determined by laboratory confirmation).

Fuel claddamage should be assumed to be the cause of elevated Reactor Coolant activity unlessanother cause is known.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category RA ICs.Basis Reference(s):

1. NEI 99-01 Rev 6, SU32. Technical Specifications 3.4.6, RCS Specific Activity3. Technical Specifications 3.7.5, Main Condenser Offgas4. DBD P-S-30, Offgas System5. ARC 218 20C210 E-1, Air Ejector Discharge Radiation High-High Month 20XXPBAPS 3-22EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FG1Initiating Condition:

Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.Operating Mode Applicability:_

1,2,3Emergency Action Level (EAL):Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.Basis:Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the General Emergency classification level each barrier is weighted equally.Basis Reference(s):

"1. NEI 99-01 Rev 6, Table 9-F-2Month 20XXPBAPS 3-23EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FS1Initiating Condition:

..... ..Loss or Potential Loss of ANY two barriers.

Operating Mode Applicability:

1,2,3Emergency Action Level (EAL): ....Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.Basis:Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the Site Area Emergency classification level, each barrier is weighted equally..Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2Month 20XXPBAPS 3-24EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FA1Initiating Condition:

ANY Loss or ANY Potential Loss of either Fuel Clad or RCS.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.Basis:Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the Alert classification level, Fuel Cladding and RCS barriers are weighted moreheavily than the Containment barrier.

Unlike the Containment

barrier, loss or potential loss of either the Fuel Cladding or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability.

Note that the loss or potential loss ofContainment barrier in combination with loss or potential loss of either Fuel Cladding orRCS barrier results in declaration of a Site Area Emergency under EAL FS1.Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2Month 20XXPBAPS 3-25EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FCiInitiating Condition:

RCS ActivityOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSA. (Site Specific.

in~dications that reactor coolant acti 'FAt' is greater than 300u~i/gm dosequivalentI activity

> 300 uCilgm Dose Equivalent 1-131.Basis:This threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gmdose equivalent 1-131. Reactor coolant activity above this level is greater than thatexpected for iodine spikes and corresponds to an approximate range of 2% to 5% fuelclad damage. Since this condition indicates that a significant amount of fuel claddamage has occurred, it represents a loss of the Fuel Clad Barrier.It is recognized that sample collection and analysis of reactor coolant with highlyelevated activity levels could require several hours to complete.

Nonetheless, asample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with RCS Activity.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2Month 20XXPBAPS 3-26EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FC2Initiating Condition:

RPV Water LevelOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSA-.1. Plant conditions indicate Primary Containment flooding is required.

iiMaFycoentainment floeodng irqurd.POTENTIAL LOSSA:-.2. RPV water level cannot be restored and maintained above (site specific RPV...ate level to the top ofce fuel)> -172 inches (TAF).Gr-OR3. RPV water level cannot be determined.

Basis:Loss 2-.AThreshold

1. 1 BasisThe Loss threshold represents the EOP requirement for primary containment flooding.

This is identified in the BWROG EPGsISAGsEOPs/SAMPs when the phrase, "PrimaryContainment Flooding Is Required,"

appears.

Since a site-specific RPV water level isnot specified here, the Loss threshold phrase, "Primary containment flooding required,"

also accommodates the EOP need to flood the primary containment when RPV waterlevel cannot be determined and core damage due to inadequate core cooling isbelieved to be occurring.

Potential Loss 2-AThreshold

1. 2 and #3 Basis:This water level corresponds to the top of the active fuel and is used in the EOPs toindicate a challenge to core cooling.The RPV water level threshold is the same as RCS balrieF-Barrier RC2 Loss -threshold 2-X Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Lossof the RCS barrier that appropriately escalates the emergency classification level to aSite Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV-wate RPV water level cannot be restored and maintained above thespecified level following depressurization of the RPV (either manually, automatically orby failure of the RCS barrier) or when procedural guidance or a lack of low pressureRPV injection sources preclude Emergency RPV depressurization.

EOPs allow theoperator a wide choice of RPV injection sources to consider when restoring RPV waterlevel to within prescribed limits. EOPs also specify depressurization of the RPV in orderMonth 20XXPBAPS 3-27EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION to facilitate RPV water level control with low-pressure injection sources.

In some events,elevated RPV pressure may prevent restoration of RPV water level until pressure dropsbelow the shutoff heads of available injection sources.

Therefore, this Fuel Clad barrierPotential Loss is met only after either: 1) the RPV has been depressurized, or requiredemergency RPV depressurization has been attempted, giving the operator anopportunity to assess the capability of low-pressure injection sources to restore RPVwater level or 2) no low pressure RPV injection systems are available, precluding RPVdepressurization in an attempt to minimize loss of RPV inventory.

The term "cannot be restored and maintained above" means the value of RPV waterlevel is not able to be brought above the specified limit (top of active fuel). Thedetermination requires an evaluation of system performance and availability in relationto the RPV water level value and trend. A threshold prescribing declaration when athreshold value cannot be restored and maintained above a specified limit does notrequire immediate action simply because the current value is below the top of activefuel, but does not permit extended operation below the limit; the threshold must beconsidered reached as soon as it is apparent that the top of active fuel cannot beattained.

Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to"restore and maintain" level above TAF resulting in this threshold being met.In high-power ATWS/failure to scram events, EOPs may direct the operator todeliberately lower RPV water level to top of ac-tive fue! in order to reduce reactorpower. RPV water lev.el is then coentrolled-boP1.A9oen the top of ac-tive fuel -and theMfinimum Stea~mCooling RP WAter level (MSR.A^L).

ýAlthough such action is achallenge to core cooling and the Fuel Clad barrier, the immediate need to reducereactor power is the higher priority.

For such events, ICs SA5-MA3 or SS,-MS3 willdictate the need for emergency classification.

Since the loss of ability to determine if adequate core cooling is being provided presentsa significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier isspecified.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. T-BAS INTRO, INTRODUCTION TO TRIPS AND SAMPS -BASES3. T-101, RPV Control Bases4. T-1 11, Level Restoration Month 20XXPBAPS 3-28EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FC5Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSA. PDi-;a-y rcontainment radiation r-eading greater than (site specific value)Drywell radiation monitor reading > 9.55 E+02 R/hr.Basis:Less 4.The radiation monitor reading corresponds to an instantaneous release of all reactorcoolant mass into the primary containment, assuming that reactor coolant activityequals 300 ltCi/gm dose equivalent 1-131. Reactor coolant activity above this level isgreater than that expected for iodine spikes and corresponds to an approximate rangeof 2% to 5% fuel clad damage. Since this condition indicates that a significant amountof fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCSBarrier RC5 Loss Tthreshold-4.A-since it indicates a loss of both the Fuel Clad Barrierand the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

There is no Fuel Clad Barrier Potential Loss threshold associated with PrimaryContainment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. Core Damage Assessment Methodology

3. ARC-005-00C226D-B-5, Unit 3 Containment Radiation Monitor Hi-Rad4. DBD P-S-43, Radiation Monitoring SystemMonth 20XXPBAPS 3-29EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION FC7Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSS1A. Any condition in the opinion of the Emergency Director that indicates Loss of theFuel Clad Barrier.POTENTIAL LOSS2A. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.Basis:Loss Threshold

1. 1 BasisThis threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.Potential Loss Threshold

1. 2 BasisPotential Lcss 6.A.This threshold addresses any other factors that may be used by the Emergency Directorin determining whether the Fuel Clad Barrier is potentially lost. The Emergency Directorshould also consider whether or not to declare the barrier potentially lost in the eventthat barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2Month 20XXPBAPS 3-30EP-AA-1007 (Revision XX)

Pa2tnh Rn*$nm Atnmir Power Stzation Annexv Exelnn Nucrleair RECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION RC2Initiating Condition:

RPV Water LevelOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSS1. RPV water level cannot be restored and maintained above (site specific RPV waterle'el .... s.......

to the top o kf tief4ue')>

-172 inches (TAF)ei-OR2. RPV water level cannot be determined.

Basis:L-ess-3This water level corresponds to the Ttop of Aactive Ffuel (TAF) and is used in the EOPsto indicate challenge to core cooling.The RPV water level threshold is the same as Fuel Clad baFBieFBarrier FC2 Potential Loss threshold-2-A.

Thus, this threshold indicates a Loss of the RCS barrier andPotential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RP-V-wat&RPV water level cannot be restored and maintained above thespecified level following depressurization of the RPV (either manually, automatically orby failure of the RCS barrier) or when procedural guidance or a lack of low pressureRPV injection sources preclude Emergency RPV depressurization EOPs allow theoperator a wide choice of RPV injection sources to consider when restoring RPV waterlevel to within prescribed limits. EOPs also specify depressurization of the RPV in orderto facilitate RPV water level control with low-pressure injection sources.

In some events,elevated RPV pressure may prevent restoration of RPV water level until pressure dropsbelow the shutoff heads of available injection sources.

Therefore, this RCS barrier Lossis met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assessthe capability of low-pressure injection sources to restore RPV water level or 2) no lowpressure RPV injection systems are available, precluding RPV depressurization in anattempt to minimize loss of RPV inventory.

The term, "cannot be restored and maintained above," means the value of RPV waterlevel is not able to be brought above the specified limit (top of active fuel). Thedetermination requires an evaluation of system performance and availability in relationto the RPV water level value and trend. A threshold prescribing declaration when aMonth 20XXPBAPS 3-31EP-AA-1007 (Revision XX)

I=xAInn Pannrh Rnffm Atnmir Power Sta~tion Anne~x Exelon Nucla~rRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION threshold value cannot be restored and maintained above a specified limit does notrequire immediate action simply because the current value is below the top of activefuel, but does not permit extended operation beyond the limit; the threshold must beconsidered reached as soon as it is apparent that the top of active fuel cannot beattained.

Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to"restore and maintain" level above TAF resulting in this threshold being met.In high-power ATWS/failure to scram events, EOPs may direct the operator todeliberately lower RPV water level to the top of acti.e fuel in order to reduce reactorpower. RP'V water l.....l *. thn con.tr.d.

beF::een the top of ac.tiv..

feand thehAMinimum.

Steam Cooling RPV Water Le.el

.)T, Although such action is achallenge to core cooling and the Fuel Clad barrier, the immediate need to reducereactor power is the higher priority.

For such events, ICs SA5-MA3 or SS&MS3 willdictate the need for emergency classification.

There is no RCS Potential Loss threshold associated with RPV Water Level.Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. T-BAS INTRO, INTRODUCTION TO TRIPS AND SAMPS -BASES3. T-101, RPV Control Bases4. T-1 11, Level Restoration Month 20XXPBAPS 3-32EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION RC3Initiating Condition:

Primary Containment PressureOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSA. Primnarc nGtainment prnessure greater than (Site Specific value) due to RCS leakage-.

1. Drywell pressure

>2.0 psig.AND2. Drywell pressure rise is due to RCS leakageBasis:Loss4-AThe (site Specific value)> 2.0 psig primary containment pressure is the d4wellDrywell high pressure setpoint which indicates a LOCA by automatically initiating equivalent makeup system.The second threshold condition focuses the fission product barrier loss threshold on afailure of the RCS instead of the non-LOCA malfunctions that may adversely affectprimary containment pressure.

Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primarycontainment vent/purge.

The release of mass from the RCS due to the as-designed/expected operation of arelief valve does not warrant an emergency classification.

A stuck-open Safety ReliefValve (SRV) or SRV leakage is not considered either identified or unidentified leakageby Technical Specification and, therefore, is not applicable to this EAL.There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. T-102, Primary Containment Control-Bases

3. T-101, RPV ControlMonth 20XXPBAPS 3-33EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION RC4Initiating Condition:

RCS Leak RateOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSAl. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU, or RCIC linebreak. en (st "i sytm with ,,,+,,,al f,, high e,, .OR82. Emergency RPV Depressurization is required.

POTENTIAL LOSS3A. UNISOLABLE primary system leakage that results in EITHER of the following:

a4-. Secondary Containment area temperature

> T-103 Alarm Setpoint.

Max NorMalOperating Temperature.

ORb2. Secondary Containment area radiation level > T-103 Alarm Setpoint.

Max-NOrmal Operating Area Radiation Level.Basis:UNISOLABLE:

An open or breached system line that cannot be isolated, remotely orlocally.Classification of a system break over system leakage is based on information available to the Control Room from the event. Indications that should be considered are:" Reports describing magnitude of steam or water release.* Use of system high flow alarms / indications, if available,

" Significant changes in makeup requirements,

" Abnormal reactor water level changes in response to the event.The use of the above indications provides the Control Room the bases to determine thatthe on going event is more significant than the indications that would be expected fromsystem leakage and therefore should be considered a system break.Month 20XXPBAPS 3-34EP-AA-1007 (Revision XX)

Poach Roftorn Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION Loss Threshold

1. 1 Basis-3ALarge high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of theRCS until they are isolated.

If it is determined that the ruptured line cannot be promptlyisolated from the Control Room, the RCS barrier Loss threshold is met.Loss Threshold

1. 2 Basis4BEmergency RPV Depressurization in accordance with the EOPs is indicative of a loss ofthe RCS barrier.

If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though theRCS is being vented into the .upr. *potorus, a Loss of the RCS barrier existsdue to the diminished effectiveness of the RCS to retain fission products within itsboundary.

Potential Loss Threshold-

1. 3 Basis 3.APotential loss of RCS based on primary system leakage outside the primarycontainment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate adirect path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expectedto occur during normal plant operating conditions with all directly associated supportand control systems functioning properly.

The indicators reaching the threshold barriers and confirmed to be caused by RCSleakage from a primary system warrant an Alert classification.

A primary system isdefined to be the pipes, valves, and other equipment which connect directly to the RPVsuch that a reduction in RPV pressure will effect a decrease in the steam or water beingdischarged through an unisolated break in the system.In general, multiple indications should be used to determine if a primary system isdischarging outside Primary Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the ReactorBuilding since this may be caused by radiation shine from nearby steam lines or themovement of radioactive materials.

Conversely, a high area radiation condition inconjunction with other indications (e.g. room flooding, high area temperatures, reports ofsteam in the Reactor Building, an unexpected rise in Feedwater

flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

An UNISOLABLE leak which is indicated by Max Normal Operating values escalates toa Site Area Emergency when combined with Containment Barrier CT6 Loss tThreshold

1. 11A (after a containment isolation) and a General Emergency when the Fuel CladBarrier criteria is also exceeded.

Month 20XXPBAPS 3-35EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. T-103, Secondary Containment ControlMonth 20XXPBAPS 3-36EP-AA-1007 (Revision XX)

I=xAInn NuclearPannrh Rnffnm Atnnmir Power Station Annex Exelnn NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION RC5Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSDrywell radiation monitor reading > 100RIhr.A. Prim~a~'

containmcnt radiation reading greater than (site sepocfic value).Basis:The radiation monitor reading corresponds to an instantaneous release of all reactorcoolant mass into the primary containment, assuming that reactor coolant activityequals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier FC5 Loss Tthreshold 4-A- since it indicates a loss of the RCSBarrier only.There is no RCS Potential Loss threshold associated with Primary Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. EP-EAL-061 1, Criteria for Choosing Containment Radiation Monitor ReadingIndicative of Loss of RCS BarrierMonth 20XXPBAPS 3-37EP-AA-1007 (Revision XX)

Pesach nttonm Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION RC7Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSAl. Any condition in the opinion of the Emergency Director that indicates Loss of theRCS Barrier.POTENTIAL LOSSA2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.Basis:Loss 6.AThreshold

1. 1 Basis:This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS Barrier is lost.Potential Loss 6-,AThreshold

1. 2 Basis:This threshold addresses any other factors that may be used by the Emergency Directorin determining whether the RCS Barrier is potentially lost. The Emergency Directorshould also consider whether or not to declare the barrier potentially lost in the eventthat barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2Month 20XXPBAPS 3-38EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION CT2Initiating Condition:

RPV Water LevelOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

POTENTIAL LOSSA7.Plant conditions indicate Primary CGontainment flooding is required.

Basis:

Loss 2. AThe Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV Water Level 2-A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and thatcore damage is possible.

BWR specify the conditions thatrequire primary containment flooding.

When primary containment flooding is required, the EPGsEOPs are exited and SAGsSAMPs are entered.

Entry into SAGsSAMPs is alogical escalation in response to the inability to restore and maintain adequate corecooling.PRA studies indicate that the condition of this Potential Loss threshold could be a coremelt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure.

In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of aGeneral Emergency.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. T-1 11, Level Restoration (LR)- Bases3. T-1 16, RPV Flooding

-Bases4. T-1 17, Level/Power Control -Bases5. TRIP/SAMP Curves, Tables, And Limits -Bases6. T-BAS (INTRO),

Introduction To Trips And Samps -BasesMonth 20XXPBAPS 3-39EP-AA-1007 (Revision XX)

Pane-h Rnffnm Afnmir- Pnuuar -Q*nfirn AnngvIPYalnn MinehlorB r.h ~n44r~m A *r~m ir~ Dr~w~r ~~~*ir~n A n n~v Fv~Inn IJuir~I~2r RECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION CT3Initiating Condition:

Primary Containment Conditions Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSAl. UNPLANNED rapid drop in.- .tainmenDrywell pressure following primarycontainment pressure rise.OR82. Primary ..

.r-ell pressure response not consistent with LOCAconditions.

POTENTIAL LOSSA3. P-imnay t pressure greater than (Site specific v..alue)>

56 psigand rising.OR134. (site specifc explosive m.iture) e.xists.

inside primay co.ntainment

a. Drywell orTorus Hydrogen concentration

> 6%.ANDb. Drywell or Torus Oxygen concentration

> 5%.ORG5. HT-LG-Heat Capacity Limit (T-102 Curve T/T-1) exceeded.

Basis:UNPLANNED:

A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.Loss I-A-and 1-BThreshold

1. 1 and #2 BasisRapid UNPLANNED loss of primary containment pressure (i.e., not attributable todFywelDrywell spray or condensation effects) following an initial pressurerise indicates a loss of primary containment integrity.

Primary containment pressure shouldi arise as a result of mass and energy release into the primary containment from aLOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for thecondition and therefore a specific value is not assigned.

The unexpected (UNPLANNED) response is important because it is the indicator for a containment Month 20XXPBAPS 3-40EP-AA-1007 (Revision XX)

I=xAInn NunlAarPp&rmh Rntnm Afnmiun Power Staution Annex ~ Exelnn NucleairRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION bypass condition.

A pressure suppression bypass path would not be an indication of acontainment breach.Potential Loss 4AThreshold

1. 3 BasisThe threshold pressure is the primary containment internal design pressure.

Structural acceptance testing demonstrates the capability of the primary containment to resistpressures greater than the internal design pressure.

A pressure of this magnitude isgreater than those expected to result from any design basis accident and, thus,represent a Potential Loss of the Containment barrier.Potential Loss 4-BThreshold

1. 4 BasisIf hydrogen concentration reaches or exceeds the lower flammability limit, as defined inplant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If thecombustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.Potential Loss 1-CThreshold

1. 5 BasisThe Heat Capacity TomperaturoF Limit (HCTLI) is the highest suppreso pootemperature from Which EmFergency RPV Depressurization Will not raise:*Suppression1 cha~mb-er temperature above the manximu~m temperature capability ofmna" be required to operate when the RPV ispeuredOR*Supeio chamber pressure above Primnary Containment Pressure Limit A,hile_ the4 rante of en~ergy transfder fro~m the RPV to the Gontaiwnmet is greater thathe capacity of the contafinment vent.The HCTL is a function of RPV pressure, suprsio potorus temperature andýý torus water level. It is utilized to preclude failure of the containment and' equipment in the containment necessary for the safe shutdown of the plant andtherefore, the inability to maintain plant parameters below the limit constitutes apotential loss of containment.

Basis Reference(s):

1 .NEI 99-01 Rev 6, Table 9-F-22. T-BAS INTRO, INTRODUCTION TO TRIPS AND SAMPS -BASES3. UFSAR Section 5.2.4, Primary Containment Safety Evaluation

4. T-102 Primary Containment Control-Bases

5. DBD P-T-12, Design Basis Accidents, Transients and Events6. DBD P-T-02, Containment, Section 3.2.14Month 20XXPBAPS 3-41EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION CT5Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

POTENTIAL LOSSA. Priarn,'

containment radiation monGitor reading grcatcr than (site apecific Value)Drywell radiation monitor reading > 2.20 E+03 R/hr.Basis:There is no Loss threshold associated with Primary Containment Radiation.

Poentwial Less 4.A.The radiation monitor reading corresponds to an instantaneous release of all reactorcoolant mass into the primary containment, assuming that 20% of the fuel cladding hasfailed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1 228, Source Estimations During Incident Response to Severe Nuclear PowerPlant Accidents, indicates the fuel clad failure must be greater than approximately 20%in order for there to be a major release of radioactivity requiring offsite protective actions.

For this condition to existT there must already have been a loss of the RCSBarrier and the Fuel Clad Barrier.

It is therefore prudent to treat this condition as apotential loss of containment which would then escalate the emergency classification level to a General Emergency.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. Core Damage Assessment Methodology

3. ARC-005-00C226D-B-5 Unit 3 Containment Radiation Monitor Hi-Rad4. DBD P-S-43, Radiation Monitoring SystemMonth 20XXPBAPS 3-42EP-AA-1007 (Revision XX)

Peacrh Bottom Atomic~ Power Station Annex Exelon Nuce~IarRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION CT6Initiating Condition:

Primary Containment Isolation FailureOperating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSAl. UNISOLABLE direct downstream pathway to the environment exists after primarycontainment isolation signal.OR82. Intentional Pprimary Ceontainment venting/purging per EOP's or SAMPs due toaccident conditions.

ORC. UNISOLABLE primary system leakage that results in EITHER of the following:

1. Secondary Containment area temperature

> T-103 Action Level. Max-SafeOperating OR2. Secondary Containment area radiation level > T-103 Action Level. Max-SafeOperating Radiation Levell.Basis:UNISOLABLE:

An open or breached system line that cannot be isolated, remotely orlocally.These thresholds address incomplete containment isolation that allows anUNISOLABLE direct release to the environment.

Loss 3AThreshold

1. 1 BasisThe use of the modifier "direct" in defining the release path discriminates againstrelease paths through interfacing liquid systems or minor release pathways, such asinstrument lines, not protected by the Primary Containment Isolation System (PCIS).Leakage into a closed system is to be considered only if the closed system is breachedand thereby creates a significant pathway to the environment.

Examples includeunisolable Main Steamline, HPCI or RCIC steamline breaks, unisolable RWCU systembreaks, and unisolable containment atmosphere vent paths.Examples of "downstream pathway to the environment" could be through theTurbine/Condenser, or direct release to the Turbine or Reactor Building.

The existence of a filter is not considered in the threshold assessment.

Filters do notremove fission product noble gases. In addition, a filter could become ineffective due toiodine and/or particulate loading beyond design limits (i.e., retention ability has beenexceeded) or water saturation from steam/high humidity in the release stream.Month 20XXPBAPS 3-43EP-AA-1007 (Revision XX)

ExAIon Pane-h Rnftnm Atnmir Power A$safinn Annex Exelon NuclanrRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION Following the leakage of RCS mass into primary containment and a rise in primarycontainment

pressure, there may be minor radiological releases associated withallowable primary containment leakage through various penetrations or systemcomponents.

Minor releases may also occur if a primary containment isolation valve(s)fails to close but the primary containment atmosphere escapes to an enclosed system.These releases do not constitute a loss or potential loss of primary containment butshould be evaluated using the Recognition Category A-RICs.Loss 38Threshold

1. 2 BasisEOPs may direct primary containment isolation valve logic(s) to be intentionally

bypassed, even if offsite radioactivity release rate limits will be exceeded.

Under theseconditions with a valid primary containment isolation signal, the containment should alsobe considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure orcombustible gas control to the secondary containment and/or the environment is a Lossof the Containment.

Venting for primary containment pressure control when not in anaccident situation (e.g., to control pressure below the dPywe!Drywell high pressurescram setpoint) does not meet the threshold condition.

Loss "GThreshold

1. 3 BasisThe Max Safe Operating Temperature and the Max Safe Operating Radiation Level areeach the highest value of these parameters at which neither:

(1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for thesafe shutdown of the plant will be precluded.

EOPs utilize these temperatures andradiation levels to establish conditions under which RPV depressurization is required.

The temperatures and radiation levels should be confirmed to be caused by RCSleakage from a primary system. A primary system is defined to be the pipes, valves, andother equipment which connect directly to the RPV such that a reduction in RPVpressure will effect a decrease in the steam or water being discharged through anunisolated break in the system.In general, multiple indications should be used to determine if a primary system isdischarging outside Primary Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the ReactorBuilding since this may be caused by radiation shine from nearby steam lines or themovement of radioactive materials.

Conversely, a high area radiation condition inconjunction with other indications (e.g. room flooding, high area temperatures, reports ofsteam in the Reactor Building, an unexpected rise in Feedwater

flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

In combination with RCS Barrier RC4 pPotential ILoss Threshold

1. 33A this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.Month 20XXPBAPS 3-44EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-22. T-1 03 Secondary Containment Control3. T-102 Primary Containment Control4. T-200-3 Primary Containment VentingMonth 20XXPBAPS 3-45EP-AA-1007 (Revision XX)

Poae-h nffnm Afnrnie-Pnuwor Afafinn AnnoyPvo-Inn konr-l,-ar Dn~ rh Rr%44r~m A*r~mir Dniuor ~*~$ir~n Annoy Fvolnn kI.ur~In~mr RECOGNITION CATEGORYFISSION PRODUCT BARRIER DEGRADATION CT7Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

1,2,3Fission Product Barrier (FPB) Threshold:

LOSSAl. Any condition in the opinion of the Emergency Director that indicates Loss of theContainment Barrier.POTENTIAL LOSSA2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containent Barrier.Basis:Loss 6-AThreshold

1. 1 Basis:This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment Barrier is lost.Potential Loss &AThreshold

1. 2 Basis:This threshold addresses any other factors that may be used by the Emergency Directorin determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in theevent that barrier status cannot be monitored.

Basis Reference(s):

1 .NEI 99-01 Rev 6, Table 9-F7-2Month 20XXPBAPS 3-46EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSG1Initiating Condition:

Prolonged loss of all Off-site and all On-Site AC power to emergency busses.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the General E- --

promptlyupon determining that (site .pecii hours) the applicable time has beenexceeded, or will likely be exceeded.

1--.7 Loss of ALL offsite "and- AL' ons,;itte AC power to unit (site 6pecifc buses)4KV safeguards Buses.AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply power tounit 4KV Safeguards Buses.AND3b. EITHER of the following:

a. Restoration of at least one unit eemegey-4KV Safeguards bus-Bus -in < 2hours is notless than (site .pecific hours) is no likely.ORb. RPV water level cannot be restored and maintained

> -195 inches.*(Site specific ind~icationA of an; inability to adequately remove heat Qfrom the core)Basis:SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

This IC addresses a prolonged loss of all power sources to AC emergency buses. Aloss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heatremoval/pressure

control, spent fuel heat removal and the ultimate heat sink. Aprolonged loss of these buses will lead to a loss of one or more fission product barriers.

Month 20XXPBAPS 3-47EP-AA-1007 (Revision XX)

Peach Boftom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS In addition, fission product barrier monitoring capabilities may be degraded under theseconditions.

The EAL should require declaration of a General Emergency prior to meeting thethresholds for IC FG1. This will allow additional time for implementation of offsiteprotective actions.Escalation of the emergency classification from Site Area Emergency will occur if it isprojected that power cannot be restored to at least one AC emergency bus by the endof the analyzed station blackout coping period. Beyond this time, plant responses andevent trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation.

Mitigation actions with a low probability of success should notbe used as a basis for delaying a classification upgrade.

The goal is to maximize thetime available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC powerresults in parameters that indicate an inability to adequately remove decay heat fromthe core.Basis Reference(s):

1. NEI 99-01 Rev 6, SG12. UFSAR Section 8.4, Auxiliary Power Systems3. UFSAR Figure 8.3.1, Transmission System4. DBD P-S-05, 4KV System5. T-101 RPV Control6. SE-11, LOSS OF OFF-SITE POWER -BASESMonth 20XXPBAPS 3-48EP-AA-1007 (Revision XX)

I=xAInn Pe~ach Rnttnm Atomic~ Power Sta~tion Annex FvaeInn Nm~uclear RECOGNITION CATEGORYSYSTEM MALFUNCTIONS MsSIInitiating Condition:

Loss of all offsite and all onsite AC power to emergency busses for 15 minutes orlonger.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the S.it;e Ar pea Emege1n;-yvent promptlyupon determining that the applicable time 1-5 miutes-has been exceeded, or willlikely be exceeded.

1. Loss of ALL offsite and on""o AC Power to (site speific emer"gency bu.es)unit 4KV Safeguards Buses. for 15 minutes or longe...AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply power to unit4KV Safeguards Buses.AND3. Failure to restore power to at least one unit 4KV Safeguards bus in < 15 minutesfrom the time of loss of both offsite and onsite AC powerBasis:SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of allSAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure

control, spent fuel heat removal andthe ultimate heat sink. In addition, fission product barrier monitoring capabilities may bedegraded under these conditions.

This IC represents a condition that involves actual orlikely major failures of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary powerlosses.Escalation of the emergency classification level would be via ICs RAG1, FG1, -e-MSG1, or MG2.Month 20XXPBAPS 3-49EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexRECOGNITION CATEGORYSYSTEM MALFUNCTIONS Basis Reference(s):

1. NEI 99-01 Rev 6, SS12. UFSAR Section 8.4, Auxiliary Power Systems3. UFSAR Figure 8.3.1, Transmission System4. DBD P-S-05, 4KV System5. SE-1 1 LOSS OF OFF-SITE POWER -BASESExelon NuclearMonth 20XXPBAPS 3-50EP-AA-1007 (Revision XX) i=xelon NuclearPezach ontnom Atomic~ Power Station Annex Exelon Nucla~rRECOGNITION CATEGORYSYSTEM MALFUNCTIONS M9A1Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the eventAled promptly upondetermining that the applicable time 4-5-minuteshas been exceeded, or will likelybe exceeded.

1. AC power capability to unit 4KV Safeguards Buses reduced to only one of thefollowing power sources for> 15 minutes.* 2 Emergency Auxiliary Transformer (OAX04)* 3 Emergency Auxiliary Transformer (OBX04)* El Diesel Generator

• E2 Diesel Generator

• E3 Diesel Generator

• E4 Diesel Generator

a. AC poweFrcapability to (site specific emergency buses) is reduced to a singlpower source for 15 minutes Or logRer.AND2b. Any additional single power source failure will result in a loss of ALLaH AC power toSAFETY SYSTEMS.Basis:SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources suchthat any additional single failure would result in a loss of all AC power to SAFETYSYSTEMS.

In this condition, the sole AC power source may be powering one, or morethan one, train of safety-related equipment.

This IC provides an escalation path from ICMsUI.An "AC power source" is a source recognized in AOPs and EOPs, and capable ofsupplying required power to an emergency bus. Some examples of this condition arepresented below.Month 20XXPBAPS 3-51EP-AA-1007 (Revision XX)

Pand-h ^++nm A+nmid- Pnwar A+afinn AnnovP=Yalnn kRir-lon=r Pn~vh Rn44r~m A*nmia- Pnwar ~*~*inn Annoy FvoIr~n hi. .r.Io~rRECOGNITION CATEGORYSYSTEM MALFUNCTIONS

" A loss of all offsite power with a concurrent failure of all but one emergency powersource (e.g., an onsite diesel generator).

" A loss of all offsite power and loss of all emergency power sources (e.g., onsitediesel generators) with a single train of emergency buses being back-fed from theunit main generator.

" A loss of emergency power sources (e.g., onsite diesel generators) with a singletrain of emergency buses being back-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary lossesof power.Escalation of the emergency classification level would be via IC MSS1.Basis Reference~s):

1. NEI 99-01 Rev 6, SA12. UFSAR Section 8.4, Auxiliary Power Systems3. UFSAR Figure 8.3.1, Transmission System4. DBD P-S-05, 4KV System5. DBD P-T-13, Station Blackout6. SE-11 LOSS OF OFF-SITE POWER -BASESMonth 20XXPBAPS 3-52EP-AA-1007 (Revision XX)

I:xelon NuclearPeach Rnttnm Atomic Power Station Annex Exelon Nucla~rRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MsUIInitiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the Unusual Event vent promptly upondetermining that the applicable time 16- m-iutes-has been exceeded, or will likelybe exceeded.

1-. Loss of ALL offsite AC power capability to unit 4KV Safeguards Buses (site-spec.ific em.ergency bu.es) for > 15 Basis:This IC addresses a prolonged loss of offsite power. The loss of offsite power sourcesrenders the plant more vulnerable to a complete loss of power to AC emergency buses.This condition represents a potential reduction in the level of safety of the plant.For emergency classification

purposes, "capability" means that an offsite AC powersource(s) is available to the emergency buses, whether or not the buses are poweredfrom it.Fifteen minutes was selected as a threshold to exclude transient or momentary lossesof offsite power.Escalation of the emergency classification level would be via IC MSA1.Basis Reference(s):

1. NEI 99-01 Rev 6, SU12. UFSAR Section 8.4, Auxiliary Power Systems3. UFSAR Figure 8.3.1, Transmission System4. DBD P-S-05, 4KV System5. DBD P-T-13, Station Blackout6. SE-11 LOSS OF OFF-SITE POWER -BASESMonth 20XXPBAPS 3-53EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexPvAninn khir-Loor RECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSG281Initiating Condition:

Loss of all AC and Vital DC power sources for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the Gen-ral Emergncy

, vent promptlyupon determining that the applicable time

-has been exceeded, or willlikely be exceeded.

1. Loss of ALL offsite AC power to unit 4KV Safeguards Buses.AND2. Failure of El, E2, E3, and E4 Emergency Diesel Generators to supply power tounit 4KV Safeguards Buses.AND3. Voltage is < 107.5 VDC on unit 125 VDC battery buses 2(3)0D021, 2(3)0D022, 2(3)0D023, and 2(3)0D024.

AND4. ALL AC and Vital DC power sources have been lost for > 15 minutes.1. a. =Lerr- onf A. I- I- ;, c- f- ,t e -;a ^1 nd-,i A I _nR t ',A ...... ; (site 6e ift ...........

-buses) for 15 mFniuters or longer-.ANIDb. Indicated voltage is less than (Site specific bus voltage value) onALbh. a g *^ .i4.. 5. I Ln r r"ir A ,~r, .., , ,Basis:SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

Month 20XXPBAPS 3-54EP-AA-1007 (Revision XX)

Pant-h Rntfnm Atomic~ Power Staition Annex ExeIon NuclanrRECOGNITION CATEGORYSYSTEM MALFUNCTIONS This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. Aloss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heatremoval/pressure

control, spent fuel heat removal and the ultimate heat sink. A loss ofVital DC power compromises the ability to monitor and control SAFETY SYSTEMS.

Asustained loss of both AC and DC power will lead to multiple challenges to fissionproduct barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary powerlosses. The 15-minute emergency declaration clock begins at the point when all EALsconditions are met.Basis Reference(s):

1. NEI 99-01 Rev 6, SG82. UFSAR Section 8.4, Auxiliary Power Systems3. UFSAR Figure 8.3.1, Transmission System4. DBD P-S-05, 4KV System5. DBD P-S-01A, 125/250 Vdc Station Batteries Including 125 Vdc Batteries

&Chargers, Table T3.2-1, MCCs, and DC Distribution

6. SE-13 LOSS OF A 125 OR 250 VDC SAFETY RELATED BUS7. ARC-220-20C209R H-3, 2B DC Power Panel Lo Voltage8. SE-11 LOSS OF OFF-SITE POWER -BASESMonth 20XXPBAPS 3-55EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSS281Initiating Condition:

Loss of all vital DC power for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the Site Area promptlyupon determining that the applicable time 45 minAes-has been exceeded, or willlikely be exceeded.

tndiGated-Voltage is < 107.5 VDC less than (site sp..ifi, bus voltage value) on 125VDC battery buses 2(3)0D021, 2(3)0D022, 2(3)0D023, and 2(3)0D024 ALL-(site-specific V.. ital 0 .busses) for >1.5 minutes-lengeBasis:SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

This IC addresses a loss of Vital DC power which compromises the ability to monitorand control SAFETY SYSTEMS.

In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary powerlosses.Escalation of the emergency classification level would be via ICs RAG1, FG1 orMSG28.Basis Reference(s):

1. NEI 99-01 Rev 6, SS82. DBD P-S-01A, 125/250 Vdc Station Batteries Including 125 Vdc Batteries

&Chargers, Table T3.2-1, MCCs, and DC Distribution

3. SE-13, LOSS OF A 125 OR 250 VDC SAFETY RELATED BUS4. ARC-220-20C209R H-3, 2B DC Power Panel Lo VoltageMonth 20XXPBAPS 3-56EP-AA-1007 (Revision XX)

NuclAarPoac~h Bottom Atomic~ Power Staition Annex Exelnn NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSS36 1Initiating Condition:

Inability to shutdown the reactor causing a challenge to RPV water level or RCS heatremoval.Operating Mode Applicability:

1,2Emergency Action Level (EAL):1. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 4%.AND2. ALLAh manual / ARI actions to shutdown the reactor have been unsuccessful asindicated by Reactor Power > 4%.AND3. EITHER of the following conditions exist:" RPV water level cannot be restored and maintained

> -195 inchesOR" Heat Capacity Limit (T-102 Curve SP/T-1) exceeded.

(Site spccific indir.ation of an inability to adequately rcmo':o heat fromn the core)(Site specific.

indication Of aniaiito adequately removo heat from the RCS)Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manualreactor scram that results in a reactor shutdown, all subsequent operator manualactions, both inside and outside the Control Room including driving in control rods andboron injection, su bseq.u.nt operator actions to .a n .ually shutdown the reactor areunsuccessful, and continued power generation is challenging the capability toadequately remove heat from the core and/or the RCS. This condition will lead to fueldamage if additional mitigation actions are unsuccessful and thus warrants thedeclaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may behigher than that resulting from an assessment of the plant responses and symptomsagainst the Recognition Category F ICs/EALs.

This is appropriate in that theRecognition Category F ICs/EALs do not address the additional threat posed by afailure to shutdown the reactor.

The inclusion of this IC and EAL ensures the timelydeclaration of a Site Area Emergency in response to prolonged failure to shutdown thereactor.Month 20XXPBAPS 3-57EP-AA-1007 (Revision XX)

Pant-h nf# m Afnmii- Pnuuar Afnfinn AnnovF::valnn Ni anr~lDn~ r~h Rn44r~m A*r~mir Pnw~r ~*2*inn Annoy Fvolnn Mu.t~Io~r RECOGNITION CATEGORYSYSTEM MALFUNCTIONS A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RAG1 or FG1.Basis Reference(s):

1. NEI 99-01 Rev 6, SS52. T-1 17, Level/Power Control3. T-101, RPV Control4. T-102, Primary Containment ControlMonth 20XXPBAPS 3-58EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSA361Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manualactions taken at the reactor control consoles are not successful in shutting down thereactor.Operating Mode Applicability:

1,2Emergency Action Level (EAL):Note: A manual action is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core, and does not include manuallydriving in control rods or implementation of boron injection strategies.

1. Af,-aAutomatic or manual scram did not shutdown the reactor as indicated byReactor Power > 4%.AND2. Manual / ARI actions taken at the reactor control conesolReactor Console are notsuccessful in shutting down the reactor as indicated by Reactor Power > 4%.Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manualreactor scram that results in a reactor shutdown, and subsequent operator manualactions taken at the reactor GOctFOI consoles to shutdown the reactor are alsounsuccessful.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if thereactor is subsequently shutdown by an action taken away from the reactor sentrteIconsoles since this event entails a significant failure of the RPS.A manual action at the reactor GentrFe consoles is any operator action, or set of actions,which causes the control rods to be rapidly inserted into the core (e.g., initiating amanual reactor scram. This action does not include manually driving in control rods orimplementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactorOntrcI consoles (e.g., locally opening breakers).

Actions taken at back-panels or otherlocations within the Control Room, or any location outside the Control Room, are notconsidered to be "at the reactor ccntrci consoles".

Taking the Reactor Mode Switch to SHUTDOWN-Shutdown is considered to be amanual scram action.The plant response to the failure of an automatic or manual reactor scram will varybased upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged Month 20XX PBAPS 3-59 EP-AA-1007 (Revision XX)

Pane-h nffem Afernie-Pewuor -Qfnfien AnnoyI=yolnn ki"Aal~r~s h Rn44r~m A *u~m ir~ Dnuu~r ~ *iir~n A n n~v Fv~Inn MImvI~2rRECOGNITION CATEGORYSYSTEM MALFUNCTIONS enough to cause a challenge to the RPV water level or RCS heat removal safetyfunctions, the emergency classification level will escalate to a Site Area Emergency viaIC MS$35. Depending upon plant responses and symptoms, escalation is also possiblevia IC FS1. Absent the plant conditions needed to meet either IC MSS36 or FS1, anAlert declaration is appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL areincluded to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Basis Reference~s):

1. NEI 99-01 Rev 6, SA52. T-1 17, Level/Power Control3. T-101, RPV ControlMonth 20XXPBAPS 3-60EP-AA-1007 (Revision XX)

Pdatxe-h nffnm Afnmie- Pnuwar Qf!2firn AnnavIPv--nn Mi~rl-tar

,. k R,%44,~m A*nmiv Dnui~r ~*~*irhn Ann~v F~i'~I,~n Mu .r~I~arRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSU361Initiating Condition:

Automatic or manual scram fails to shutdown the reactor.Operating Mode Applicability:

1,2Emergency Action Level (EAL):Note: A manual action is any operator action, or set of actions, which causes thecontrol rods to be rapidly inserted into the core, and does not include manuallydriving in control rods or implementation of boron injection strategies.

1. a. AR-aAutomatic scram did not shutdown the reactor as indicated by ReactorPower > 4%..ANDb. A-sSubsequent manual / ARI action taken at the reactor cOnGeO selesReactor Console is successful in shutting down the reactor.OR2. a. A-mManual scram -did not shutdown the reactor as indicated by ReactorPower > 4%.ANDb. EITHER of the following:

1. A-sSubsequent manual / ARI action taken at the reaGcGor-cntrol Console is successful in shutting down the reactor.OR2. A-sSubsequent automatic

/ ARI -scram is successful in shutting downthe reactor.Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manualreactor scram that results in a reactor shutdown, and either a subsequent operatormanual action taken at the reactor GGen#-.consoles or an automatic scram is successful Iin shutting down the reactor.

This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.EAL #1 BasisFollowing the failure on an automatic reactor scram, operators will promptly initiatemanual actions at the reactor G0i.-.c!

onsoles to shutdown the reactor (e.g., initiate amanual reactor scram). If these manual actions are successful in shutting down thereactor, core heat generation will quickly fall to a level within the capabilities of theplant's decay heat removal systems.Month 20XXPBAPS 3-61EP-AA-1007 (Revision XX)

Peach Boftom Atomic Power Station AnnexExellon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS EAL #2 BasisIf an initial manual reactor trip is unsuccessful, operators will promptly take manualaction at another location(s) on the reactor GoAtrel-consoles to shutdown the reactor(e.g., initiate a manual reactor scram/ARI using a different switch).

Depending uponseveral factors, the initial or subsequent effort to manually scram the reactor, or aconcurrent plant condition, may lead to the generation of an automatic reactor scramsignal. If a subsequent manual or automatic scram/ARI is successful in shutting downthe reactor, core heat generation will quickly fall to a level within the capabilities of theplant's decay heat removal systems.A manual action at the reactor GontroI consoles is any operator action, or set of actions,which causes the control rods to be rapidly inserted into the core (e.g., initiating amanual reactor scram). This action does not include manually driving in control rods orimplementation of boron injection strategies.

Actions taken at back-panels or otherlocations within the Control Room, or any location outside the Control Room, are notconsidered to be "at the reactor citr9I-consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scramaction.The plant response to the failure of an automatic or manual reactor tscram will varybased upon several factors including the reactor power level prior to the event,availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at thereactor Gentr-eIconsoles are also unsuccessful in shutting down the reactor, then theemergency classification level will escalate to an Alert via IC MSA35. Depending uponthe plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC MSA35 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpointtesting),

the following classification guidance should be applied." If the signal generated as a result of plant work causes a plant transient that createsa real condition that should have included an automatic reactor scram and the RPSfails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

" If the signal generated as a result of plant work does not cause a plant transient butshould have generated an RPS scram signal and the scram failure is determined through other means (e.g., assessment of test results),

then this IC and the EALsare not applicable and no classification is warranted.

Month 20XXPBAPS 3-62EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexRECOGNITION CATEGORYSYSTEM MALFUNCTIONS Basis Reference(s):

1. NEI 99-01 Rev 6, SU52. T-1 17, Level/Power ControlExelon NuclearMonth 20XXPBAPS 3-63EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSA42Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with asignificant transient in progress.

Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the eventAJert promptly upondetermining that the applicable time 15-inutes-has been exceeded, or will likelybe exceeded.

1. a. AR--UNPLANNED event results in the inability to monitor one Or meoANYTable Ml f the fe...Wi, g parameters from within the Control Room for >15minutes GF-IegR.[seetable b~elw]!BWR parameter listjReac~tor Power___________________

Table M1 Control Room Parameters RP7 Water

• Reactor Power, RPV Water LevelRPV Prcssuc RPV Pressure* Drywell PressurePrimary Containment

• Torus LevelPresure , Torus Temperature Supp.ession Pool LevelSuppression Pool Temperattur-e ANDb. Any Table M2ef he transient eyeits in progress.

" Automatic.

Or Manual runback gratc than 25% thermal reactor power" Elecr-tric-al load rejoction; greater than 2624 fu-ll elcrclelad

" Reactkfp" EGGS (Sb) actuation Month 20XXPBAPS 3-64EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS Table M2 Significant Transients

• Automatic or Manual Runback >25% thermal reactor power" Electrical Load Rejection

>25% full electrical load" Reactor Scram" ECCS Actuation

" Thermal Power oscillations

> 10%Basis:UNPLANNED:

A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring rapidly changing plantconditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fissionproduct barrier challenge is reduced.

It thus represents a potential substantial degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of thelisted parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

Forexample, the reactor power level cannot be determined from any analog, computerpoint, digital and recorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems isevaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1 022)to determine if an NRC event report is required.

The event would be reported if itsignificantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective actiondecision-making.

This EAL is focused on a selected subset of plant parameters associated with the keysafety functions of reactivity

control, RI-V-levelRPV water level and RCS heat removal.Month 20XXPBAPS 3-65EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS The loss of the ability to determine one or more of these parameters from within theControl Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for one or more of the listed parameters are lost,then the ability to determine the values of other SAFETY SYSTEM parameters may beimpacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary lossesof indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RASI.Basis Reference(s):

1. NEI 99-01 Rev 6, SA2Month 20XXPBAPS 3-66EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station Annex M,,&lAmrRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSU42Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the 'JUua! E'ventevent promptly upondetermining that the applicable time 45-Fiiute6-has been exceeded, or will likelybe exceeded.

a-. AR-UNPLANNED event results in the inability to monitor one or- morANY Table M1parameters from within the Control Room for_> 15 minutes.Table M1 Control Room Parameters

• Reactor Power* RPV Water Level* RPV Pressure* Drywell Pressure* Torus Level* Torus Temperature

1. of thel fol lh4Rq paramt.re fro m within the Control Ro m for 15 mninuter.,

2. IRW-R prammef 3. IPWR paramctcr

4. Reactor-Power- 6. Reacter-Pewer8. RPV Water- L"vl9RG-ee

10. R.PV Pressure

11. RCS P".essufe 1-2. Pimary 13.1n Corc'Come ceijoininemo Exit Tc--p.,-ture Pressure14. Suppression Pool 15. Levels in at leastLeve (site speeifinumber)stam

16. Suppression Poo 17. S;teafm GCencritor TempeFatur-e Axle-___Water lowMonth 20XXPBAPS 3-67EP-AA-1007 (Revision XX)

I:xAInn NuclAarPoaneh Rnffnm Afnmivr Pnwar S$fation Annex Exelnn Nucla~rRECOGNITION CATEGORYSYSTEM MALFUNCTIONS Basis:UNPLANNED:

A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the ControlRoom. This condition is a precursor to a more significant event and represents apotential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of thelisted parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

Forexample, the reactor power level cannot be determined from any analog, digital andrecorder source within the Control Room.An event involving a loss of plant indications, annunciators and/or display systems isevaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1 022)to determine if an NRC event report is required.

The event would be reported if itsignificantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective actiondecision-making.

This EAL is focused on a selected subset of plant parameters associated with the keysafety functions of reactivity

control, core cooling and RCS heat removal.

The loss ofthe ability to determine one or more of these parameters from within the Control Roomis considered to be more significant than simply a reportable condition.

In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well.For example, if the value for reactor vessel level cannot be determined from theindications and recorders on a main control board, the SPDS or the plant computer, theavailability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary lossesof indication.

Escalation of the emergency classification level would be via IC MSA42-.Month 20XXPBAPS 3-68EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexRECOGNITION CATEGORYSYSTEM MALFUNCTIONS Basis Reference(s):........

1. NEI 99-01 Rev 6, SU2Exelon NuclearMonth 20XXPBAPS 3-69EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSA59Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed-required for the currentoperating mode.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):(-1) 1. a- The occurrence of ANY of the following hazardous events:" Seismic event (earthquake)

" Internal or external flooding event" High winds or tornado strike" FIRE" EXPLOSIONSpecific hazards)* Other events with similar hazard characteristics as determined by theShift ManagerAND2.b-. EITHER of the following:

a.4-1 Event damage has caused indications of degraded performance inat least one train of a SAFETY SYSTEM ipeeded-required byTechnical Specifications for the current operating mode.ORb.2L. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEMcomponent or structure needed-required by Technical Specifications for the current operating mode.Basis:FIRE: Combustion characterized by heat and light. Sources of smoke such as slippingdrive belts or overheated electrical equipment do not constitute FIRES. Observation offlame is preferred but is NOT required if large quantities of smoke and heat areobserved.

EXPLOSION:

A rapid, violent and catastrophic failure of a piece of equipment due tocombustion, chemical reaction or overpressurization.

A release of steam (from highenergy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Suchevents may require a post-event inspection to determine if the attributes of an explosion are present.Month 20XXPBAPS 3-70EP-AA-1007 (Revision XX)

Peach Boftorn Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS SAFETY SYSTEM: A system required for safe plant operation, cooling down the plantand/or placing it in the cold shutdown condition, including the ECCS. These aretypically systems classified as safety-related.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements,

testing, or analysis.

The visual impact of the damage issufficient to cause concern regarding the operability or reliability of the affectedcomponent or structure.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or astructure containing SAFETY SYSTEM components, needed-required for the currentoperating mode, "required",

i.e. required to be operable by Technical Specifications forthe current operating mode. This condition significantly reduces the margin to a loss orpotential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. Manual or automatic electrical isolation of safety equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.EAL 144#2.a BasisThis EALaddresses damage to a SAFETY SYSTEM train that is required to beoperable by Technical Specifications for the current operating mode, and is inseFViGeloperation since indications for it will be readily available.

The indications ofdegraded performance should be significant enough to cause concern regarding theoperability or reliability of the SAFETY SYSTEM train.EAL 1.b22.b BasisThis EAL addresses damage to a SAFETY SYSTEM component that is required to beoperable by Technical Specifications for the current operating mode, and is not insefviGeloperation or readily apparent through indications alone, er-as well as damage toa structure containing SAFETY SYSTEM components.

Operators will make thisdetermination based on the totality of available event and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level would be via IC FS1 or RAS1.If the EAL conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.Basis Reference(s):

1. NEI 99-01, Rev 6 SA92. SE-5, Earthquake

-Bases3. SEA, Flood -BasesMonth 20XXPBAPS 3-71EP-AA-1007 (Revision XX)

FxAInn rPp Vt-h Rnttnm Avnmire Pnwar vtvinn Annn v iimlvn NuciearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS

4. DBD P-T-09, Internal HazardsMonth 20XXPBAPS 3-72EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearPeach Bottom Atomic Power Station Annex Exelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSU641Initiating Condition:

RCS leakage for 15 minutes or longer.Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):Note: The Emergency Director should declare the Unusual Event vent promptly upondetermining that the applicable time 15-inutes-has been exceeded, or will likelybe exceeded.

1. RCS unidentified or pressure boundary leakage in the Drywell greate--than

> 10 gpm for > 15 minutes.

(site specific value) for 15 minutes or longer.OR2. RCS identified leakage in the Drywell gFeater-than->25 gpm for > 15 minutes.(site-specific value) for 15 mninutes or longer.OR3. Leakage from the RCS to a location outside the DrywellGentai-meRt

>25 gpm for >15 minutes ge. ater than 25 gpm for 15 mninutes or longe.Basis:UNISOLABLE:

An open or breached system line that cannot be isolated, remotely orlocally.This IC addresses RCS leakage which may be a precursor to a more significant event.In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.EAL #1 and EAL #2 BasisThese EALs are focused on a loss of mass from the RCS due to "unidentified leakage",

"pressure boundary leakage" or "identified leakage" (as these leakage types are definedin the plant Technical Specifications).

EAL #3 BasisThis EAL addresses a RCS mass loss caused by an UNISOLABLE leak through aninterfacing system.These EALs thus apply to leakage into the containment, a secondary-side systemsteam generator tube leakage in a PWR) or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.

Lesser values typically require time-consuming Month 20XXPBAPS 3-73EP-AA-1007 (Revision XX)

I=xelon NuclearPesach onttnm Atomic~ Power Station Annex Exelon NuclanrRECOGNITION CATEGORYSYSTEM MALFUNCTIONS calculations to determine (e.g., a mass balance calculation).

EAL #1 uses a lower valuethat reflects the greater significance of unidentified or pressure boundary leakage.The release of mass from the RCS due to the as-designed/expected operation of arelief valve does not warrant an emergency classification.

Fr -BWR'VR,

-Aa stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified orunidentified leakage by Technical Specifications and, therefore, is not applicable to thisEAL.The 15-minute threshold duration allows sufficient time for prompt operator actions toisolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category RA or F.Basis Reference(s):

1. NEI 99-01 Rev 6, SU42. Technical Specifications 3.4.4, RCS Operational Leakage3. UFSAR 4.10.3, Nuclear System Leakage Detection and Leakage Rate Limits4. UFSAR 5.2.5, Primary Containment

-Inspection and Testing5. ST-O-020-560-2/3, Reactor Coolant Leakage Test6. DBD P-S-34 Radwaste System, 3.3.1.3.1 Dfywe!!Drywell Floor Drain SumpOperation

7. DBD P-S-34 Radwaste System, 3.3.1.3.2 gFywe'lDrywell Equipment Drain SumpOperation Month 20XXPBAPS 3-74EP-AA-1007 (Revision XX)

Peach Bottom Atomic Power Station AnnexExelon NuclearRECOGNITION CATEGORYSYSTEM MALFUNCTIONS MSU761Initiating Condition:

Loss of all On-site or Off-site communications capabilities.

Operating Mode Applicability:

1,2,3Emergency Action Level (EAL):1. Loss of ALL Table M3 Onsite communications capability affecting the ability toperform routine operations.

OR2. Loss of ALL Table M3 Offsite communication capability affecting the ability toperform offsite notifications.

OR3. Loss of ALL Table M3 NRC communication capability affecting the ability toperform NRC notifications.

Table M3 Communications Capability System Onsite Offsite NRCStation Radio System XPlant Public Address (PA) XOMNI System X XStation Phones X X XSatellite Phones X X XNARS XHPN X XENS X Xal II1. 6066 O? ALL OT Me rolloWIna onsite communIcation monoas:(Site Weneific lost of Gemmunircafians mnethod* .... .iv ......................

--\ .... ZL -- J--SL~ec or !~LL r TflO ronowia tn-~cmuijin imro_zP_1wI ...........

(i~t-e 6s9-9Ocif lIBt Of communications methods)4 Month 20XXPBAPS 3-75EP-AA-1007 (Revision XX)

Pa2r-h nffnm Afnmie Pnuwar A#2finn AnnayFwainn Mid-loarP~2r~h RnI$nm Atnmir~ Pnw~r ~t2tinn A nn~v Fv~lnn Mmuv~I~2r RECOGNITION CATEGORYSYSTEM MALFUNCTIONS

3. Los f ALL- o-f the follow~ing NRC communications mM-etos(soite specific-list ofcomnatnsmhd)

Basis:This IC addresses a significant loss of on-site,

-ef-offsite, or NRC communications capabilities.

While not a direct challenge to plant or personnel safety, this eventwarrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.This IC should be assessed only when extraordinary means are being utilized to makecommunications possible (e.g., use of non-plant, privately owned equipment, relaying ofon-site information via individuals or multiple radio transmission points, individuals beingsent to offsite locations, etc.).EAL #1 BasisaAddresses a total loss of the communications methods used in support of routine plantoperations.

EAL #2 BasisaAddresses a total loss of the communications methods used to notify all OROs of anemergency declaration.

The OROs referred to here are listed in procedure EP-MA-1 14-100-F-01, State/Local Event Notification Form. (see DveepeF Notes).EAL #3 BasisaAddresses a total loss of the communications methods used to notify the NRC of anemergency declaration.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU62. UFSAR Section 10.21, Communication Systems3. EP-MA-124-1001 Facility Inventories and Equipment Tests4. NE-CG-400-19 Lighting and Communication DrawingsMonth 20XXPBAPS 3-76EP-AA-1007 (Revision XX)