ML14164A076

Attachment 10: Discussion of Revision to the Radiological Emergency Plan Annex for Quad Cities Nuclear Power Station, Cover Through Page QC 3-68

ML14164A076
Person / Time
Site:	Quad Cities
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 ML14164A067 ML14164A068 ML14164A069 ML14164A071 ML14164A073 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 9: Discussion of Revision to the Radiological Emergency Plan Annex for Peach Bottom Atomic Power Station, Page PBAPS 3-77 Through End 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-1006, NEI 99-01, Rev. 6
Download: ML14164A076 (136)

Text

ATTACHMENT 10 DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FOR QUAD CITIES NUCLEAR POWER STATION EP-AA-1 006 Enclosures 0 Enclosure 10A - EAL Comparison Matrix Document 0 Enclosure 10B - EAL Red-Line Basis Document S Enclosure 10C - EAL Basis Document

NEI 99-01 REVISION 6 DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS ATTACHMENT 10 DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FOR QUAD CITIES NUCLEAR STATION m Exelon Generation.

NEI 99-01 Rev 6 Proposed EAL Justification AG1 RG1 Initiating Condition - GENERAL EMERGENCY Initiating Condition: 1'1 No Change M Difference FIDeviation Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem Release of gaseous radioactivity resulting in offsite dose greater than 1,000 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 5,000 mrem thyroid CDE. mRem TEDE or 5,000 mRem thyroid CDE. classification.

Operating Mode Applicability: All Operating Mode Applicability:

2) Added the following to bullet #3" Classification based on effluent monitor 1,2.3,4,5,D readings assumes that a release path to the environment is established." In order Emergency Action Level (EAL): to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) readings assumes that a release path to the environment is established. If the Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for

• The Emergency Director should declare the General Emergency promptly classification purposes." This allows for more timely classification since all the upon determining that the applicable time has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded.

that the applicable time has been exceeded, or will likely be exceeded. readily available on the 11x17 procedure matrix used by the SM.

• If an ongoing release is detected and the release start time is unknown,

• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes. assume that the release duration has exceeded 15 minutes.

• If the effluent flow past an effluent monitor is known to have stopped due to

• Classification based on effluent monitor readings assumes that a release path actions to isolate the release path, then the effluent monitor reading is no to the environment is established. If the effluent flow past an effluent monitor is longer valid for classification purposes. known to have stopped due to actions to isolate the release path, then the

• The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.

used for emergency classification assessments until the results from a dose

• The pre-calculated effluent monitor values presented in EAL #1 should be assessment using actual meteorology are available.

used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer:

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCilsec for > 15 minutes (as determined by Control Room Panels or PPDS

- Total Noble Gas Release Rate).

(site specific monitor list and threshold values)

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the

2. Dose assessment actual meteorology indicates doses greater than 1000 site boundary of EITHER:

mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose receptor point) a. > 1000 mRem TEDE OR

3. Field survey results indicate EITHER of the following at or beyond (site b. > 5000 mRem CDE Thyroid specific dose receptor point): OR

• Closed window dose rates greater than 1000 mR/hr expected to 3. Field survey results at or beyond the site boundary indicate EITHER:

continue for 60 minutes or longer.

a. Gamma (closed window) dose rates >1000 mR/hr are expected to

• Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.

5000 mrem for one hour of inhalation.

OR

b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

Page I of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification AS1 RS1 Initiating Condition - SITE AREA EMERGENCY Initiating Condition: D No Change I Difference 1-- Deviation Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem Release of gaseous radioactivity resulting in offsite dose greater than 100 mRem 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 500 mrem thyroid CDE. TEDE or 500 mRem thyroid CDE. classification.

Operating Mode Applicability: All Operating Mode Applicability:

2) Added the following to bullet #3" Classification based on effluent monitor 1,2,3,4,5,D readings assumes that a release path to the environment is established." In order Emergency Action Level (EAL): to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) readings assumes that a release path to the environment is established. If the Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for

• The Emergency Director should declare the Site Area Emergency promptly classification purposes." This allows for more timely classification since all the upon determining that the applicable time has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded. that the applicable time has been exceeded, or will likely be exceeded. readily available on the 1 1x17 procedure matrix used by the SM.

• If an ongoing release is detected and the release start time is unknown, I an ongoing release is detected and the release start time is unknown, If assume that the release duration has exceeded 15 minutes.

assume that the release duration has exceeded 15 minutes.

• If the effluent flow past an effluent monitor is known to have stopped due to

• Classification based on effluent monitor readings assumes that a release path actions to isolate the release path, then the effluent monitor reading is no to the environment is established. If the effluent flow past an effluent monitor is longer valid for classification purposes.

known to have stopped due to actions to isolate the release path, then the

• The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.

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

for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer:

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS -

Total Noble Gas Release Rate).

(site specific monitor list and threshold values)

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the

2. Dose assessment actual meteorology indicates doses greater than 1000 site boundary of EITHER:

mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose receptor point) a. > 100 mRem TEDE OR

3. Field survey results indicate EITHER of the following at or beyond (site b. > 500 mRem CDE Thyroid specific dose receptor point):

OR

• Closed window dose rates greater than 100 mR/hr expected to 3. Field survey results at or beyond the site boundary indicate EITHER:

continue for 60 minutes or longer.

a. Gamma (closed window) dose rates >100 mR/hr are expected to

" Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.

500 mrem for one hour of inhalation.

OR

b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.

Page 2 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification AA1 RA1 Initiating Condition - ALERT Initiating Condition:

H No Change Difference L- Deviation Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 1) Listed site-specific monitors and Threshold values to ensure timely classification.

mrem TEDE or 50 mrem thyroid CDE. mrem TEDE or 50 mrem thyroid CDE.

Operating Mode Applicability: All Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1,2,3,4, 5, D to delete the following from the basis "Classification based on effluent monitor Emergency Action Level (EAL): readings assumes that a release path to the environment is established. Ifthe Example Emergency Action Levels: (1 or 2 or 3) effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." This allows for more timely classification since all the basis

• The Emergency Director should declare the Alert promptly upon information pertaining to Note bullet 3 will be contained in the IC and therefor determining that the applicable time has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining readily available on the 11x17 procedure matrix used by the SM.

exceeded. 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,

• If an ongoing release is detected and the release start time is unknown, assume 3) A calculation was performed, in accordance with (IAW) guidance provided in NEI assume that the release duration has exceeded 15 minutes. 99-01 revision 6 EAL AA1, to determine the effluent monitor response for a that the release duration has exceeded 15 minutes.

radioactive liquid release via the normal site release pathway. The liquid release

• If the effluent flow past an effluent monitor is known to have stopped due would contain activity equivalent to provide 10mrem TEDE or 50mrem thyroid CDE Classification based on effluent monitor readings assumes that a release path to actions to isolate the release path, then the effluent monitor reading is to the environment is established. If the effluent flow past an effluent monitor is at the site boundary. The calculation determined the effluent monitor response no longer valid for classification purposes. known to have stopped due to actions to isolate the release path, then the would be >110% of the instrument maximum range and as such, lAW NEI 99-01

• The pre-calculated effluent monitor values presented in EAt #1 should be effluent monitor reading is no longer valid for classification purposes. Rev 6 guidance, was not included in this EAL.

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

1. Reading on any of the following radiation monitors greater than the

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 reading shown for 15 minutes or longer: uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS

- Total Noble Gas Release Rate).

OR (site-specific monitor list and threshold values)

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

2. Dose assessment actual meteorology indicates doses greater than 10

a. > 10 mRem mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose TEDE receptor point)

OR

3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem b. > 50 mRem CDE Thyroid thyroid CDE at or beyond (site-specific dose receptor point) for one hour of OR exposure.

3. Analysis of a liquid effluent sample indicates a concentration or release rate

4. Field survey results indicate EITHER of the following at or beyond (site that would result in doses greater than EITHER of the following at or beyond specific dose receptor point): the site boundary

• Closed window dose rates greater than 10 mR/hr expected to a. 10 mRem TEDE for 60 minutes of exposure continue for 60 minutes or longer.

OR

• Analysis of field survey samples indicate thyroid CDE greater than b. 50 mRem CDE Thyroid for 60 minutes of exposure 50 mrem for one hour of inhalation.

OR

4. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates > 10 mRihr are expected to continue fork 60 minutes.

OR

b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.

Page 3 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification AU1 RUI Initiating Condition - UNUSUAL EVENT Initiating Condition: 1 No Change E Differencee Deviation Release of gaseous or liquid radioactivity greater than 2 times the (site-specific Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 1) Listed site-specific monitors and Threshold values to ensure timely classification.

effluent release controlling document) limits for 60 minutes or longer 60 minutes or longer.

Operating Mode Applicability: Alt Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1,2,3,4,5, D to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." This allows for more timely classification since all the basis

• The Emergency Director should declare the Unusual Event promptly upon

• The Emergency Director should declare the event promptly upon determining information pertaining to Note bullet 3 will be contained in the IC and therefor determining that 60 minutes has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded.

readily available on the 11x17 procedure matrix used by the SM.

exceeded.

• If an ongoing release is detected and the release start time is unknown, assume

• If an ongoing release is detected and the release start time is unknown, that the release duration has exceeded 60 minutes.

assume that the release duration has exceeded 60 minutes.

• Classification based on effluent monitor readings assumes that a release path

• If the effluent flow past an effluent monitor is known to have stopped, to the environment is established. If the effluent flow past an effluent monitor is indicating that the release path is isolated, the effluent monitor reading is known to have stopped due to actions to isolate the release path, then the no longer valid for classification purposes. effluent monitor reading is no longer valid for classification purposes.

1. Reading on ANY effluent radiation monitor greater than 2 times the (site- 1. Reading on ANY of the following effluent monitors > 2 times alarm specific effluent release controlling document) limits for 60 minutes or setpoint established by a current radioactive release discharge permit longer: for a 60 minutes.

(site-specific monitor list and threshold values corresponding to 2 times

• Radwaste Effluent Monitor 1/2-1799-01 the controlling document limits)

OR

• Discharge Permit specified monitor

2. Reading on ANY effluent radiation monitor greater than 2 times the OR alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCilsec for > 60 minutes (as determined by Control Room Panels or PPDS

- Total Noble Gas Release Rate).

3. Sample analysis for a gaseous or liquid release indicates a concentration OR or release rate greater than 2 times (site-specific effluent release controlling document limits) for 60 minutes or longer. 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.

Page 4 of 66

NEI 99-01 Rev 6 Proposed EAL Justification AG2 RG2 No Change Difference Deviation Initiating Condition - GENERAL EMERGENCY Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer. 1) EAL not used in accordance with the discussion in Section 1.4, NRC Order EA-1 2-051, it is recommended that this EAL be implemented when the enhanced Operating Mode Applicability: All spent fuel pool level instrumentation is available for use. The completion of the enhanced SFP level indicators and need for the inclusion of this EAL is being Example Emergency Action Levels: tracked in accordance with Exelon Generation Company, LLC's Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for NOTES: The Emergency Director should declare the General Emergency Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated promptly upon determining that 60 minutes has been exceeded, October 25,2012.

or will likely be exceeded

1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer.

Page 5 of 66

NEI 99-01 Rev 6 Proposed EAL Justification AS2 RS2 Initiating Condition - SITE AREA EMERGENCY AR No Change 1X- Difference []Deviation Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) 1) EAL not used in accordance with the discussion in Section 1.4, NRC Order EA-1 2-051, it is recommended that this EAL be implemented when the enhanced Operating Mode Applicability: All spent fuel pool level instrumentation is available for use. The completion of the enhanced SFP level indicators and need for the inclusion of this EAL is being Example Emergency Action Levels: tracked in accordance with Exelon Generation Company, LLC's Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for

1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated description) October 25,2012.

Page 6 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition - ALERT I Condition: RA2 D No Change E Difference F-- 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: Ail Operating Mode Applicability:

2) EAL #3 not used in accordance with the discussion in Section 1.4, NRC Order 1, 2, 3, 4, 5, D EA-1 2-051, it is recommended that this EAL be implemented when the enhanced Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): spent fuel pool level instrumentation is available for use. The completion of the enhanced SFP level indicators and need for the inclusion of this EAL is being

1. 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 Status OR Report to March 12, 2012 Commission Order Modifying Licenses with Regard for

2. Damage to irradiated fuel resulting in a release of radioactivity from the Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated fuel 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 R1 Radiation Monitor reading >1000 mRem/hr.

(site-specific listing of radiation monitors, and the associated readings, Table R1 setpoints and/or alarms) Fuel Handling Incident Radiation Monitors

3. Lowering of spent fuel pool level to (site-specific Level 2 value).

• 1(2) 1705-16A Fuel Pool Rad Monitor

. 1(2) 1705-16B Fuel Pool Rad Monitor Page 7 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification AU2 RU2 Initiating Condition: UNUSUAL EVENT Initiating Condition: E No Change 1 Difference I: Deviation UNPLANNED loss of water level above irradiated fuel. UNPLANNED loss of water level above irradiated fuel 1) Listed site specific level indication and monitors to ensure timely classification.

Operating Mode Applicability: All Operating Mode Applicability:

1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

1. a. UNPLANNED water level drop in the REFUELING PATHWAY as 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by indicated by ANY of the following: ANY of the following:

" Refueling Cavity water level < 282 in. (Upper Wide range (site-specific level indications).

simulated signal).

OR AND

" Spent Fuel Pool water level < 19 ft. above the fuel (< -4 ft.

indicated level).

OR

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors. " Indication or report of a drop in water level in the REFUELING PATHWAY.

AND (site-specific list of area radiation monitors)

b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R1.

Table RI Fuel Handling Incident Radiation Monitors

• 1(2) 1705-16A Fuel Pool Rad Mon

• 1(2) 1705-16B Fuel Pool Rad Mon Page 8 of 66

Rev 66 55-01 Rev Proposed EAL NEI 99-01 NEI Proposed EAL Justification AA3 RA3 Initiating Condition - ALERT Initiating Condition: No Change 1 Difference II7 Deviation Radiation levels that impede access to equipment necessary for normal plant Radiation levels that impede access to equipment necessary for normal plant operations, 1) Listed site specific plant rooms and areas with identified mode applicability to operations, cooldown or shutdown. cooldown or shutdown. ensure timely classification.

Operating Mode Applicability: All Operating Mode Applicability:

1,2,3,4, 5,D Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

Note: If the equipment in the listed room or area was already inoperable, or Note:

out of service, before the event occurred, then no emergency If the equipment in the room or area listed in Table R3 was already inoperable, or classification is warranted out of service, before the event occurred, then no emergency classification is

1. Dose rate greater than 15 mR/hr in ANY of the following areas: warranted.

• Control Room

• Central Alarm Station 1. Dose rate> 15 mR/hr in ANY of the following Table R2 areas:

• (other site-specific areas/rooms) Table R2 Areas Requiring Continuous Occupancy

2. An UNPLANNED event results in radiation levels that prevent or

• Main Control Room (Unit 1 ARM Station #22) significantly impede access to any of the following plant rooms or

• Central Alarm Station - (by survey) areas:

(site-specific list of plant rooms or areas with entry-related mode OR applicability identified)

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:

Table R3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building* Mode 3 and 4 Turbine Building* Mode 3

• Areas required to establish shutdown cooling Page 9 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU3 Initiating Condition: RU3 [Z No Change Difference Deviation Initiating Condition: UNUSUAL EVENT 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 to Power Operation, Startup, Hot Standby, Hot Shutdown 1.2, 3 maintain consistency with current and previous revisions of Exelon EALs. This will Example Emergency Action Levels: Emergency Action Level (EAL): ensure a timely classification since the threshold values are more aligned with the radiological category vice system category.

1. (Site-specific radiation monitor) reading greater than (site-specific value). 1. Offgas system radiation monitor HI-HI alarm.

OR OR

2. Sample analysis indicates that a reactor coolant activity value is greater than 2. Specific coolant activity > 4.0 uCil/gm Dose equivalent 1-131.

an allowable limit specified in Technical Specifications.

Page 10 of 66

NEI 99-01 rev 6 Fission Product Barrier Matrix FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss { L Potential Loss Loss L Potential Loss 4Loss Potential Loss A. UNPLANNED rapid drop in A. Primary containment pressure

• RCS ActivityI primary containment pressure Primary greater than (site-specific value)

A. (Site specific indications that following primary containment Containment A. Primary containment pressure OR reactor coolant activity is None pressure rise Pressure / greater than (site-specific None B. (site-specific explosive mixture) greater than 300 uCilgm OR Primary value) due to RCS leakage. exists inside primary containment dose equivalent 1-131) B. Primary containment pressure Containment OR Conditions response not consistent with LOCA conditions. C. HCTL exceeded.

cannot be A. RPV water level cannot be RPV water A. restored andlevel maintained above A. Primary containment flooding restored and maintained above

2. RPV Water required. (site-specific RPV water level (site-specific RPV water level None None A. Primary containment flooding Level corresponding to top of active corresponding to the top of required.

fuel) or cannot be determined. active fuel) or cannot be determined.

A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR primary system A. UNISOLABLE break in any of the A. UNISOLABLE fUollowing: ((britepecfic following: site-specificanysthems systems leakage that EITHER of theresults in exceeding folwn:venting B. Intentional per primary EOPs containment 3.RCS Leak with potential for high-energy linefollowing:

Rate/ Primary None None breas) 1. Max Normal Operating OR None Containment OR Temperature C. UNISOLABLE primary system Isolation Failure B. Emergency RPV OR leakage that results in exceeding

2. Max Normal Operating Area EITHER of the following:

Depressurization Radiation Level. 1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

4.Primary A. Primary Containment A. Primary Containment Radiation A. Primary Containment Radiation Containment Radiation Monitor reading None Monitor reading greater than None None Monitor reading greater than (site-Radiation greater than (site-specific (site-specific value). specific value).

value).

5. Emergency A. Any Condition in the opinion A. Any Condition in the opinion of 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 Director of the Emergency Director the Emergency Director that the Emergency Director that E g Directr ta indicatestha inicate Emergency Director that indicates dgment that indicates Loss of the indicates Potential Loss of the indicates Loss of the RCS Potential Loss of the RCS BarrierP Loss of the Containment Barrier.

Judgment Fuel Clad Barrier. Fuel Clad Barrier. Barrier. P Barrier.

Page t1 of 66

Proposed Fission Product Barrier Matrix FGI Loss of any two barriers AND Loss or Potential Loss of third barrier. TED I FSI Loss or Potential Loss of ANY two barriers. IFAI ANY Loss or ANY Potential Loss of either Fuel Clad or RCS fFli]

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity> 300 uCllgm Dose None None

1. RCS Activity Equivalent 1-131. None None None

2. RPV water level cannot be restored and 1. RPV water level cannot be restored and

2. RPV Water 1. Plant conditions indicate Primary maintained 0 -142 inches (TAF) maintained > -142 Inches (TAF) Plant conditions indicate Primary Containment Containment fending is required. OR OR None None flooding is required.

Level 3. RPV water levelcannot be determined. 2. RPV water level cannot be determined.

3. Primary Containment pressure > 56 peig.

OR

1. UNPLANNED rapid drop in Drywell 4. a. Drywellor torus hydrogen concentration

1. Drywellpressure >2.5 peig. pressure followingprimary containment 06%.

3. Primary AND pressure rise. AND Containment None None 2. Drywellpressure rise is due to RCS None OR b. Drywellor tows oxygen concentration Pressure/Conditions leakage 2. Drywellpressure response not consistent a 5%.

withLOCAconditions. OR

5. Heat Capacity Limit(lGA200. Figure M) exceeded.

3. UNISOLABLE primarysystem leakage that

1. UNISOLABLE Main Steam Line(MSL), results in EITHERofthe following:

HPCI,Feedwater, RWCU or RCICline a. Secondary Containment area break. temperature s QGA300 Maximum 4.RCS Leak Rate None None OR Normal None None Emergency RPV Depressurization is

2. required. OR operating levels.

b. Secondary Containment level> OGA area radiation 300 Maximum Normal operating level.

5.Primary Drywellradiation monitor reading Drywellradiation monitor reading Drywellradiation monitor reading Containment > 6.65 E*02R/hr. None I 1O0R/hr (10.00 E+02 R/hr). None None > 1.55 E+03Rbhr Radiation

1. UNISOLABLE direct downstream pathway to the environment exists after pnmary containment isolation signal.

OR

2. Intentional Primary Containment venting/purging per EOPs or SAMGsdue to accident conditions.

6.Primary OR Containment None None None None 3. UNISOLABLE primary system leakage that None Isolation Failure results in EITHERof the following:

a. Secondary Containment area temperature s QGA300, Maximum Safe operating levels.

OR

b. Secondary Containment area radiation level s QGA300, Maximum Safe operating levels

7. Emergency A. AnyCondition in the opinion of the A.AnyCondition in the opinion of the A.ANYCondition in the opinion of the A.Any Condition in the opinion of the A.Any Condition in the opinion of the A.AnyCondition in the opinion of the Emergency Director that indicates Loss Emergency Director that indicates Potential Emergency Director that indicates Loss of Emergency Directorthat indicates Potential Emergency Directorthat indicates Loss ot the Emergency Director that indicates Potential Loss Director Judgment of the Fuel Clad Barrier. Loss ofthe Fuel Clad Barrier the RCS Barrier. Loss of the RCS Barrier. Containment Barrier. of the Containment Barrier.

Page 12 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FCI Category: Fuel Clad Barrier No Change Difference 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. (Site specific indications that reactor coolant activity is greater than 300 uCi/gm Coolant activity > 300 uCilgm Dose Equivalent 1-131.

dose equivalent 1-131)

Page 13 of 66

NEI 99-01 Rev 6 Proposed EAL Justification FC2 Category: Fuel Clad Barrier FC2 Z No Change Differce Deviation Category: Fuel Clad Barrier 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary containment flooding required. 1. Plant conditions indicate Primary Containment flooding is required.

Potential Loss Potential Loss A. RPV water level cannot be restored and maintained above (site-specific RPV 2. RPV water level cannot be restored and maintained > -142 inches (TAF) water level corresponding to top of active fuel) or cannot be determined. OR

3. RPV water level cannot be determined.

Page 14 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC4 Category: Fuel Clad Barrier 'C5 [fl 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 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 6.65 E+02 R/hr.

value).

Page 15 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC6 Category: Fuel Clad BarrierFC No Change Differ Deviation Emergency Director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. 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 of the Fuel Clad Barrier. the Fuel Clad Barrier.

Potential Loss Potential Loss A. 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

NEI 99-01 Rev 6 Proposed EAL Justification Category: Reactor Coolant System Barrier RC1 Category: Reactor Coolant System Barrier RC' E 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary containment pressure greater than (site-specific value) due to RCS 1. Drywell pressure >2.5 psig.

leakage. AND

2. Drywell pressure rise is due to RCS leakage.

Page 17 of 66

NEI 99-01 Rev 6 Proposed EAL Justification RC2 RC2 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier No Change 1 Difference 1 Deviation RPV Water Level RPV Water Level 1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A.RPV water level cannot be restored and maintained above (site-specific RPV 1. RPV water level cannot be restored and maintained > -142 inches (TAF).

level corresponding to the top of active fuel) or cannot be determined. OR

2. RPV water level cannot be determined.

Page 18 of 66

NEI 99-01 Rev 6 Proposed EAL Justification RC3 Category: Reactor Coolant System Barrier RC E No Change Difference Deviation Category: Reactor Coolant System Barrier 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. UNISOLABLE break in any of the following: ( site-specific systems with potential 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater,, RWCU or RCIC for high-energy line breas) line break.

OR OR B. Emergency RPV Depressurization 2. Emergency RPV Depressurization is required.

Potential Loss Potential Loss A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following: 3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature > QGA 300 Maximum Normal

1. Max Normal Operating Temperature. operating levels.

OR OR

2. Max Normal Operating Area Radiation Level. b. operating Containment area radiation level > QGA 300 Maximum Normal Secondarylevels.

Page 19 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Reactor Coolant System Barrier RC4 Category: Reactor Coolant System Barrier 'C5 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 100 R/hr (>1.OOE+02 R/hr).

value).

Page 20 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification RC6 RC7 NOCag [ Dfeee ]Deito Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier No Change Difference Deviation Emergency Director Judgment Emergency director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1, 2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. 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 of the RCS Barrier. the RCS Barrier.

Potential Loss Potential Loss A. 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 of66

NEI 99-01 Rev 6 Proposed EAL Justification CTI Category: Containment Barrier No Change Diff n Deviation Category: Containment Barrier 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 in Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 pressure where pressure is now lowering, the risk of a potential loss of containment is no longer present, this wording is also consistent with present EAL Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: wording.

Loss Loss C. UNPLANNED rapid drop in primary containment pressure following primary 1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.

containment pressure rise OR OR 2. Drywell pressure response not consistent with LOCA conditions.

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

3. Drywall pressure >56 psig and rising.

D. Primary containment pressure greater than (site-specific value) OR OR 4. a. Drywell or torus Hydrogen concentration > 6%.

E. (site-specific explosive mixture) exists inside primary containment AND OR b. Drywell or torus Oxygen concentration > 5%.

OR

3. HCTL exceeded. 5. Heat Capacity Limit (QGA 200, Fig. M) exceeded.

Page 22 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT2 Category: Containment Barrier CT2 No Change Difference Deviation Category: Containment Barrier RPV Water Level RPV Water Level Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss - Potential Loss A. Primary containment flooding required. Plant conditions indicate Primary Containment flooding is required.

Page 23 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT3 CT6 Category: Containment Barrier Category: Containment Barrier E No Change 1 Difference 1 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 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss LoSS A. UNISOLABLE direct downstream pathway to the environment exists after 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal primary containment isolation signal.

OR OR

2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to B. Intentional primary containment venting per EOPs accident conditions.

OR OR

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

C. UNISOLABLE primary system leakage that results in exceeding EITHER of the

a. Secondary Containment area temperature > OGA 300 Maximum Safe following:

operating levels.

1. Max Safe Operating Temperature. OR OR b. Secondary Containment area radiation level > QGA 300 Maximum Safe operating levels.

2. Max Safe Operating Area Radiation Level.

Page 24 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Containment Barrier CT4 Category: Containment BarrierT No Change Differnc 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss Potential Loss A. Primary Containment Radiation Monitor reading greater than (site-specific value). Dryweil radiation monitor reading > 1.55 E+03 R/h.

Page 25 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT6 Category: Containment BarrierCTI No Change Diffrenc. Deviation Category: Containment Barrier Emergency director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. 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 of the Containment Barrier. the Containment Barrier.

Potential Loss Potential Loss A. 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

NEI 99-01 Rev 6 j Proposed EAL Justification SG1 MG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition: F No Change FIDifference FIDeviation Prolonged loss of all offsite and all onsite AC power to emergency buses. Prolonged loss of all offsite and all onsite AC power to emergency buses.

1) Listed site specific equipment, site specific time based on station blackout Operating Mode Applicability: Operating Mode Applicability: coping analysis, and site specific indication to ensure timely classification.

1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emergency promptly upon determining that (site-specific hours) has been exceeded, or will Note:

likely be exceeded.

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

1. a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).

1 Loss of ALL offsite AC power to unit ECCS busses.

AND AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel

b. EITHER of the following: generators to supply power to unit ECCS busses.

AND

" Restoration of at least one emergency bus in less than (site-specific hours) is not likely. 3. EITHER of the following:

a. Restoration of at least one unit ECCS bus in <1 hour is not likely.

" (Site-specific indication of an inability to adequately OR remove heat from the core)

b. RPV water level cannot be restored and maintained > -166 inches.

Page 27 of 66

0 NEI 99-01 Rev 6 Proposed EAL [ Justification SS1 MS1 Initiating Condition: SITE AREA EMERGENCY Initiating Condition:

No Change FIDifference F Deviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer.

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

Operating Mode Applicability: Operating Mode Applicability:

1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Emergency Action Level (EAL):

Note: The Emergency Director should declare the Site Area Emergency Note:

promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) exceeded.

for 15 minutes or longer.

1. Loss of ALL offsite AC Power to unit ECCS buses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Page 28 of 66

NEI 99-01 Rev 6 Proposed EAL I Justification SA1 MAI Initiating Condition: ALERT Initiating Condition:

M No Change -- Difference D 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 longer.

1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

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

The Emergency Director should declare the event promptly upon

1. a. AC power capability to (site-specific emergency buses) is reduced to a determining that the applicable time has been exceeded, or will likely be single power source for 15 minutes or longer. exceeded.

AND

1. AC power capability to unit ECCS buses reduced to only one of the following

b. Any additional single power source failure will result in loss of all AC power sources for> 15 minutes.

power to SAFETY SYSTEMS.

• Reserve auxiliary Transformer TR-12 (TR-22)

• Unit Auxiliary Transformer TR-1 1 (TR-21)

• Unit Emergency Diesel Generator

• Shared Emergency Diesel Generator

• Station Blackout Diesel Generator

• Unit crosstie breakers AND

2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Page 29 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification Sul MU1 Initiating Condition: UNUSUAL EVENT Loss of all offsite AC power capability to emergency buses for 15 minutes or Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or M No Change FIDifference FIDeviation longer. longer. 1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

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

exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 exceeded.

minutes or longer Loss of ALL offsite AC power capability to unit ECCS buses for > 15 minutes.

Page 30 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SG8 Initiating Condition: GENERAL EMERGENCY Loss of all AC and Vital DC power sources for 15 minutes or longer.

Initiating Condition:

MG2 D No Change M Difference FIDeviation Loss of all AC and Vital DC power sources for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability: 2) in EAL #3 removed the word "indicated" this will allow for an indication problem Operating Mode Applicability: to not cause confusion on the need to declare.

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be Note:

exceeded.

1. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be buses) for 15 minutes or longer. exceeded.

AND

1. Loss of ALL offsite AC power to unit ECCS busses.

Indicated voltage is less than (site-specific bus voltage value) on ALL AND (site-specific vital DC buses) for 15 minutes or longer.

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.

AND

3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.

AND

4. ALL AC and Vital DC power sources have been lost for >15 minutes.

Page 31 of 66

NEt 99-01 Rev 6 Proposed EAL I Justification S33 MS2 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: D No Change M Difference I Deviation Loss of all Vital DC power for 15 minutes or longer.

Loss of all Vital DC power for 15 minutes or longer.

1) Listed site specific equipment and value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability: 2) Removed the word "indicated" this will allow for an indication problem to not Power Operation, Startup, Hot Standby, Hot Shutdown cause confusion on the need to declare.

1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:

Indicated voltage is less than (site-specific bus voltage value) on ALL Vital DC The Emergency Director should declare the event promptly upon buses for 15 minutes or longer. determining that the applicable time has been exceeded, or will likely be exceeded.

Voltage is < 105 VDC on 125 VDC battery buses #1 and #2 for >15 minutes..

Page 32 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification SS5 Initiating Condition: SITE AREA EMERGENCY Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPV Initiating Condition:

MS3 D1 No Change M Difference

• Deviation water level [BWR]) or RCS heat removal. Inability to shutdown the reactor causing a challenge to RPV water level or RCS.

heat removal. 1) Listed site specific indications to ensure timely classification.

Operating Mode Applicability: 2) Mode 2 included in operating mode applicability as per developer notes.

Operating Mode Applicability:

Power Operation 1,2 Example Emergency Action Levels:

Emergency Action Level (EAL):

1 a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the reactor.

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power

> 5%.

AND AND

b. All manual actions to shutdown the reactor have been unsuccessful. 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%'

AND AND

c. EITHER of the following conditions exist: 3. EITHER of the following conditions exist:

1. (Site-specific indication of an inability to adequately " RPV water level cannot be restored and maintained > -166 inches.

remove heat from the core) OR

" Heat Capacity Limit (QGA 200, Figure M) exceeded.

OR

2. (Site-specific indication of an inability to adequately remove heat from the RCS)

Page 33 of 66

0 NEI 99-01 Rev 6 1 Proposed EAL Justification SAO MA3 Initiating Condition: ALERT Initiating Condition: FH1 No Change M Difference FIDeviation Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and Automatic or manual scram fails to shutdown the reactor, and subsequent manual 1) Listed site specific indications to ensure timely classification.

subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. actions taken at the reactor control consoles are not successful in shutting down the reactor. 2) Mode 2 included in operating mode applicability as per developer notes.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation 1,2 Example Emergency Action Levels:

Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually Note:

driving in control rods or implementation of boron injection strategies.

A manual action is any operator action, or set of actions, which causes the

1. a. An automatic (trip [PWR] / scram [BWRJ) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor. manually driving in control rods or implementation of boron injection strategies.

AND

1. Automatic or manual scram did not shutdown the reactor as indicated by

b. Manual action taken at the reactor control consoles are not Reactor Power > 5%.

successful in shutting down the reactor. AND

2. Manual I ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.

Page 34 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU5 MU3 Initiating Condition: UNUSUAL EVENT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor.

Initiating Condition: FH No Change J--M Difference '- Deviation Automatic or manual scram fails to shutdown the reactor.

1) Listed site specific indications to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

2) Mode 2 included in operating mode applicability as per developer notes.

Power Operation 1,2 Example Emergency Action Levels: (1 or 2)

Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note:

A manual action is any operator action, or set of actions, which causes the

1. a. An automatic (trip [PWR] I scram [BWRJ) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor. manually driving in control rods or implementation of boron injection strategies.

AND

1. a. Automatic scram did not shutdown the reactor as indicated by Reactor

b. A subsequent manual action taken at the reactor control consoles is Power > 5%

successful in shutting down the reactor.

AND

b. Subsequent manual / ARI action taken at the Reactor Console is

2. a. A manual scram ([PWR] / scram [BWR]) did not shutdown the successful in shutting down the reactor.

reactor.

OR AND 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power > 5%.

b. EITHER of the following: AND

b. EITHER of the following:

1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. 1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.

OR OR

2. Subsequent automatic scram / ARI is successful in shutting down

2. A subsequent automatic (trip [PWR] / scram [BWR]) is successful the reactor.

in shutting down the reactor.

Page 35 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SA2 MA4 Initiating Condition: ALERT Initiating Condition: F No Change IDifference FI Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Note: The Emergency Director should declare the Alert promptly upon Emergency Action Level (EAL):

determining that 15 minutes has been exceeded, or will likely be exceeded.

Note:

The Emergency Director should declare the event promptly upon

1. a. An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.

[see table below]

1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter

[ BVVR parameter list] [PWR parameter list] from within the Control Room for >15 minutes.

Reactor Power Reactor Power Table M1 Control Room Parameters RPV Level RCS Level

" Reactor Power

• RPV Water Level RPV Pressure RCS Pressure

• RPV Pressure

" Drywell Pressure Primary Containment Pressure In Core/Core Exit Temperature " Torus Level

" Torus Temperature Levels in at least (site specific Suppression Pool Level number) steam generators AND 1 nPool Temperature Steam Generator Auxiliary or

2. ANY Table M2 transient in progress.

Suppression PEmergency Feed Water Flow AND Table M2 Significant Transients

b. Any of the following transient events in progress. " Turbine Trip

• Automatic or Manual runback greater than 25% thermal reactor power " Reactor Scram

• Electrical load rejection greater than 25% full electrical load " ECCS Activation

• Reactor Scram [BWR] / trip [PWR]

• ECCS (SI) actuation " Recirc. Runback > 25% Reactor Power Change

• Thermal power oscillations greater than 10% [BWR]

• Thermal Power oscillations > 10% Reactor Power Change Page 36 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU2 MU4 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change FIDifference FIDeviation UNPLANNED loss of Control Room indications for 15 minutes or longer.

UNPLANNED loss of Control Room indications for 15 minutes or longer.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Note: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):

determining that 15 minutes has been exceeded, or will likely be Note:

exceeded.

The Emergency Director should declare the event promptly upon An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.

[see table below] UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for > 15 minutes.

[ BWVRparameter list] [PWR parameter list]

Table M1 Control Room Parameters Reactor Power Reactor Power

• Reactor Power RPV Level RCS Level

• RPV Water Level

• RPV Pressure

• Drywell Pressure RPV Pressure RCS Pressure

• Torus Level

• Torus Temperature Primary Containment Pressure In Core/Core Exit Temperature Levels in at least (site specific Suppression Pool Level number) steam generators Steam Generator Auxiliary or Suppression Pool Temperature Emergency Feed Water Flow Page 37 of 66

NEI 99-01 Rev 6 j Proposed EAL Justification SA9 MA5 Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM needed for the current operating Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM required for the current operating FH No Change FI Difference H Deviation

1) No additional site specific hazard noted mode. mode.

2) Changed the word "needed" to "required" in the IC and "required by Technical Operating Mode Applicability: Operating Mode Applicability: Specification" in the EAL to be consistent with terminology used by operators and minimize confusion.

1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Level (EAL):

Example Emergency Action Levels:

1. The occurrence of ANY of the following hazardous events:

1. a. The occurrence of ANY of the following hazardous events: " Seismic event (earthquake)

• Internal or external flooding event

• Seismic event (earthquake)

• Internal or external flooding event

• High winds or tornado strike

• High winds or tornado strike

• FIRE

• FIRE " EXPLOSION

• EXPLOSION

" Other events with similar hazard characteristics as determined

• (site-specific hazards) by the Shift Manager

• Other events with similar hazard characteristics as determined by the Shift Manager AND

2. EITHER of the following:

AND a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM

b. EITHER of the following: required by Technical Specifications for the current operating mode.

1. Event damage has caused indications of degraded performance OR in at least one train of a SAFETY SYSTEM needed for the current operating mode. b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical OR Specifications for the current operating mode.

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

Page 38 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU4 MU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change M Difference FIDeviation RCS leakage for 15 minutes or longer.

RCS leakage for 15 minutes or longer. 1) Listed site specific values to ensure timely classification.

Operating Mode Applicability: 2) Changed wording from containment to Drywell for clarity to better define the Operating Mode Applicability: primary containment structure.

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 3) In EAL #1 and 2 added "into the Drywell" to differentiate between EAL #1/2 and

1. 3. Without this wording would have been in EAL #1 or #2 concurrent with #3. With Example Emergency Action Levels: (1 or 2 or 3) the added wording each EAL can be called separately.

Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unusual Event promptly upon Note:

determining that 15 minutes has been exceeded, or will likely be exceeded.

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

1. RCS unidentified or pressure boundary leakage greater than (site-specific exceeded.

value) for 15 minutes or longer.

1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for

2. RCS identified leakage greater than (site-specific value) for 15 minutes or > 15 minutes.

longer OR

2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.

3. Leakage from the RCS to a location outside containment greater than 25 OR gpm for 15 minutes or longer

3. Leakage from the RCS to a location outside the Drywell >25 gpm for

> 15 minutes.

Page 39 of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification sub MU7 Initiating Condition: UNUSUAL EVENT Loss of all onsite or offsite communications capabilities Initiating Condition: M No Change FIDifference FIDeviation Loss of all onsite or offsite communication capabilities. 1) Listed site specific communication methods to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels: (1 or 2 or 3)

Emergency Action Level (EAL):

1. Loss of ALL of the following onsite communication methods:

(site-specific list of communications method 1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine operations.

2. Loss of ALL of the following ORO communications s) methods: OR

2. Loss of ALL Table M3 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.

OR

3. Loss of ALL of the following NRC communications methods:

3. Loss of ALL Table M3 NRC communication capability affecting the ability to perform NRC notifications.

(site-specific list of communications methods)

Table M3 Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X Sound Powered Phones X All telephone Lines (Commercial and X X X microwave)

ENS X X HPN X X Satellite Phones X X Page 40 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA2 CAI Initiating Condition: ALERT Initiating Condition: E No Change 1 1 Difference [: Deviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.

longer.

Operating Mode Applicability:

Operating Mode Applicability:

4,5, D Cold Shutdown, Refueling, Defueled Emergency Action Level (EAL):

Example Emergency Action Levels:

Note:

Note: The Emergency Director should declare the Alert promptly upon The Emergency Director should declare the event promptly upon determining determining that 15 minutes time has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded.

exceeded.

Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 minutes or longer. 1. Loss of ALL offsite AC power to unit ECCS buses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Page 41 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: C No Change Difference Deviation Initiating Condition: UNUSUAL EVENT 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.

longer.

Operating Mode Applicability:

Cold Shutdown, Refueling, Defueled Operating Mode Applicability:

4, 5, D Example Emergency Action Levels:

Note: 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 single power source for 15 minutes or longer, determining that the applicable time has been exceeded, or will likely be exceeded.

AND

b. Any additional single power source failure will result in loss of all AC 1. AC power capability to unit ECCS buses reduced to only one of the power to SAFETY SYSTEMS. following power sources for > 15 minutes.

• Reserve auxiliary Transformer TR-12 (TR-22)

• Unit auxiliary transformer TR-1 1 (TR-21)

• Unit Emergency Diesel Generator

• Shared Emergency Diesel Generator

• Station Blackout Diesel Generator

• Unit crosstie breakers AND

2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Page 42 of 66

Justification Proposed EAL Rev 6 99-01 Rev NEI 99-01 6 Proposed EAL Justification Initiating Condition - ALERT Hazardous event affecting SAFETY SYSTEM needed for the current operating CA6 Initiating Condition:

Hazardous event affecting SAFETY SYSTEM required for the current operating CA2 H No Change 1 -lDifference Deviation

1) No additional site specific hazards noted mode. mode.

Operating Mode Applicability: Operating Mode Applicability:

2) Changed the word "needed" to "required" in the IC and "required by Technical Specification" in the EAL to be consistent with terminology used by Cold Shutdown, Refueling 4,5 operators and minimize confusion.

Example Emergency Action Levels: Emergency Action Level (EAL):

1. a. The occurrence of ANY of the following hazardous events: 1. The occurrence of ANY of the following hazardous events:

• Seismic event (earthquake)

• Seismic event (earthquake)

• Internal or external flooding event

• Internal or external flooding event

• High winds or tornado strike

• High winds or tornado strike

• FIRE

• FIRE

" EXPLOSION

• EXPLOSION

" (site-specific hazards)

• Other events with similar hazard characteristics as determined by the Shift Manager

" Other events with similar hazard characteristics as determined by the Shift Manager AND AND 2. EITHER of the following:

b. EITHER of the following: a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM

1. Event damage has caused indications of degraded required by Technical Specifications for the current performance in at least one train of a SAFETY SYSTEM operating mode.

needed for the current operating mode.

OR OR

b. The event has caused VISIBLE DAMAGE to a SAFETY

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current SYSTEM component or structure required by technical operating mode. Specifications for the current operating mode.

Page 43 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CU4 CU3 D N hne D Dfeec - eito Initiating 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 not Cold 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 exceeded.

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Indicated voltage is less than (site-specific bus voltage value) on required Vital DC exceeded.

buses for 15 minutes or longer.

Voltage is < 105 VDC on required 125 VDC battery buses #1 and #2 for> 15 minutes.

Page 44 of 66

NEI 99-01 Rev 6 ] Proposed EAL Justification CU5 CU4 Initiating Condition: UNUSUAL EVENT Initiating Condition: 1 -l No Change [: Difference -- Deviation Loss of all onsite or offsite communications capabilities Loss of all onsite or offsite communication capabilities. 1) Listed site specific communications methods to ensure timely classification Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown, Refueling, Defuled 4,5, D Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL):

1. Loss of ALL of the following onsite communication methods: I. Loss of ALL Table C1 Onsite communications capability affecting the ability to perform routine operations.

(site-specific list of communications method OR

2. Loss of ALL of the following ORO communications s) methods:

2. Loss of ALL Table C1 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.

3. Loss of ALL of the following NRC communications methods:

OR (site-specific list of communications methods)

3. Loss of ALL Table C1 NRC communication capability affecting the ability to perform NRC notifications.

Table Cl Communications Capab lity System Onsite Offsite NRC Plant Radio X Plant Page X Sound Powered Phones x All telephone Lines (Commercial and X X X microwave)

ENS x x HPN x x Satellite Phones X X Page 45 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA3 CAS Initiating Condition: ALERT Initiating Condition: M No Change E: Difference 1 Deviation Inability to maintain the plant in cold shutdown. Inability to maintain plant in cold shutdown. 1) Listed site specific Technical Specification cold shutdown temperature limit and site-specific pressure reading to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown, Refueling 4,5 Example 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 exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be

1. UNPLANNED increase in RCS temperature to greater than (site-specific exceeded.

Technical Specification cold shutdown temperature limit) for greater than the duration specified in the following table.

2. UNPLANNED RCS pressure increase greater than (site-specific pressure 1. UNPLANNED rise in RCS temperature > 212?F due to loss of decay reading). (This EAL does not apply during water-solid plant conditions. heat removal for > Table C2 duration.

[PWRJ)

Table C2 RCS Heat-up Duration Thresholds Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration RCS Status Containment Closure Heat-up Duration Status Status Intact Not Applicable 60 minutes*

Intact (but not RCS Reduced Not Applicable 60 minutes* Established 20 minutes*

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

reduced inventory Not Established 0 minutes

[PWR]) Not Established 0 minutes Ifan RCS heat removal system is in operation within this time

• If an RCS heat removal system is in operation within this time frame and frame and RCS temperature is being reduced, then EAL #1 is not RCS temperature is being reduced, the EAL is not applicable. asolicable.

OR

2. UNPLANNED RPV pressure rise >10 psig as a result of temperature rise due to loss of decay heat removal.

Page 46 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification CU3 CUS E Ne Initiating Condition: UNUSUAL EVENT Initiating Condition: Li No Change Difference Deviation UNPLANNED increase in RCS temperature. UNPLANNED rise in RCS temperature. 1) Listed site specific Technical Specification cold shutdown temperature limit Operating 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 likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be

1. 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 vesselIRCS [PAR] or RPV

[BWR8 ) level indication for 15 minutes or longer,. UNPLANNED rise in RCS temperature> 212F due to loss of decay heat removal.

OR

2. Loss of the following for >15 minutes.

• ALL RCS temperature indications AND

• ALL RPV water level indications Page 47 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: GENERAL EMERGENCY CGI Initiating Condition:

CG6

- No Change Wl Difference -- Deviation Loss of (reactor vessel/RCS (PWRJ or RPV [BWR]) inventory affecting fuel clad Loss of reactor vessel / RCS inventory affecting fuel clad integrity with integrity with containment challenged. containment challenged. 1) Listed site specific levels, radiation monitors, and sumps and tanks to ensure Operating Mode Applicability: timely classification.

Operating Mode Applicability:

Cold Shutdown, Refueling 4,5 2) Listed Explosive mixture in the Containment Challenge Table to ensure timely classification.

Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emergency Note: 3) Worded "cannot be monitored" as unknown to ensure clarity for instances promptly upon determining that 30 minutes time has been exceeded, or when the indicator is working but is over/under ranged. This is also in keeping The Emergency Director should declare the event promptly upon will likely be exceeded. with current EAL wording.

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

1. a. (Reactor vessel/RCS (PWR] or RPV [BWR]) vessel level less than (site- exceeded.

specific level) for 30 minutes or longer.

1. a. RPV water level < -142 inches (TAF) for > 30 minutes.

AND AND

b. ANY indication from the Containment Challenge Table b. Any Containment Challenge Indication (Table C4)

2. a.. (Reactor vessel/RCS EPWR) or RPV [SWR]) vessel level cannot be OR monitored for 30 minutes or longer. 2. a. RPV water level unknown for > 30 minutes.

AND AND

b. Core uncovery is indicated by ANY of the following: b. Core uncovery is indicated by ANY of the following:

• Table C3 indications of a sufficient magnitude to indicate core

• (Site-specific radiation monitor) reading greater than (site-specific value) uncovery.

• Erratic source range monitor indication [PWVR] OR

• UNPLANNED increase in (site-specific sump and/or tank levels) of

• Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.

sufficient magnitude to indicate core uncovery AND

• (Other site-specific indications) c. ANY Containment Challenge Indication (Table C4)

AND Table C3 Indications of RCS Leakage

c. ANY indication from the Containment Challenge Table).

• UNPLANNED floor or equipment sump level rise*

Table: Containment Challenge Table

• UNPLANNED Torus level rise*

• CONTAINMENT CLOSURE not established*

• (Explosive mixture) exists inside containment

• UNPLANNED vessel make up rate rise

• UNPLANNED increase in containment pressure

• Observation of leakage or inventory loss

• Secondary containment radiation monitor reading above (site-specific value) [BWR] 'Rise in level is attributed to a loss of RPV inventory.

• if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required. Table C4 Containment Challenge Indications

• Primary Containment Hydrogen Concentration _>6% and Oxygen

> 5%

• UNPLANNED rise in containment pressure

• CONTAINMENT CLOSURE not established' ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.

'if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.

Page 48 of 66

Justification Proposed EAL NEI 99-01 NEI Rev 66 99-01 Rev Proposed EAL Justification i

CS1 CS6 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: [-1 No Change W Difference E ] Deviation Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting core Loss of reactor vessel / RCS inventory affecting core decay heat removal 1) Listed site specific values for level, radiation monitors, and sumps and tanks decay heat removal capability. capabilities. to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability: 2) EAL #1 Quad Cities Station does not have a low, low, low RPV water level Cold Shutdown, Refueling 4,5 value stated within their Technical Specifications, all ESF functions come from the low, low value of -55.2inches, NEI 99-01 rev 6 developer notes for CS1 Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): request the developer to use the low, low, low value for declaration of a SAE Note: The Emergency Director should declare the Site Area Emergency Note: when Containment Closure is not established, as such an RPV water level -65 promptly upon determining that 30 minutes time has been exceeded, or inches was selected as the threshold value since itis substantially above TAF The Emergency Director should declare the event promptly upon RPV water level of -142 inches and substantially below the low, low value of will likely be exceeded.

determining that the applicable time has been exceeded, or will likely be -55.2inches clearly indicating a worsening of the condition and the need to

1. a. CONTAINMENT CLOSURE not established. exceeded. escalate the emergency from an Alert condition.

AND

. With CONTAINMENT CLOSURE not established, RPV water level 3) Worded "cannot be monitored" as unknown to ensure clarity for instances

b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site- when the indicator is working but is over/under ranged. This is also in keeping specific level). <-65 inches with current EAL wording.

2. a. CONTAINMENT CLOSURE established. OR AND 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 inches (TAF)

b. (Reactor vessel/RCS [PIVR] or RPV [BWVRJ)level less than (site-specific level). OR

3. a. (Reactor vessel/RCS [PWI/R] or RPV [BVVR]) level cannot be 3. a. RPV water level unknown for > 30 minutes monitored for 30 minutes or longer.

AND AND

b. Core uncovery is indicated by ANY of the following:

b. Core uncovery is indicated by ANY of the following:

" Table C3 indications of a sufficient magnitude to indicate core

• (Site-specific radiation monitor) reading greater than (site- uncovery.

specific value)

OR

• Erratic source range monitor indication [PWR]

" Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.

• UNPLANNED increase in (site-specific sump and/or tank levels) of sufficient magnitude to indicate core uncovery

• (Other site-specific indications) 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 49 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA1 CA6 Initiating Condition: ALERT Initiating Condition: D No Change W Difference Deviation Loss of (reactor vessel/RCS [PWRI or RPV [BWVRI)inventory Loss of RPV inventory 1) Listed site specific levels, and sumps and tanks to ensure timely Operating Mode Applicability: Operating Mode Applicability: classification.

Cold Shutdown, Refueling 4,5 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 keeping Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

with current EAL wording.

Note: The Emergency Director should declare the Alert promptly upon Note:

determining that 15 minutes time has been exceeded, or will likely be exceeded.

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

1. Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as exceeded.

indicated by level less than (site-specific level).

2. a. (Reactor vesselIRCS [PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer

1. Loss of RPV inventory as indicated by level < - 59inches.

AND OR

b. UNPLANNED increase in (site-specific sump and/or tank) levels due 2. a. RPV water level unknown for> 15 minutes.

to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. AND

b. Loss of RPV inventory per Table C3 indications.

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

NEI 99-01 Rev 6 Proposed EAL Justification Cu1 CU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change E Difference 1: Deviation UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory for UNPLANNED loss of reactor vessel / RCS inventory for 15 minutes or longer. 1) Described 'a required lower limit" as a procedurally established lower limit, 15 minutes or longer.

Operating Mode Applicability:

and listed site specific sumps and tanks to ensure timely classification.

Operating Mode Applicability:

4,5 Cold Shutdown, Refueling 2) Worded "cannot be monitored" as unknown to ensure clarity for instances Emergency Action Level (EAL):

Example Emergency Action Levels: (1 or 2) when the indicator is working but is over/under ranged. This is also in keeping Note: with current EAL wording.

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

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

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

[PWR] or RPV [BWRI) level less than a required lower limit for 15 minutes or longer. 1. UNPLANNED loss of reactor coolant results in the inability to restore and

2. a. (Reactor vessel/RCS [PAR] or RPV [BWVR])level cannot be maintain RPV water level above the procedurally established lower limit for> 16 minutes.

monitored.

AND OR

b. UNPLANNED increase in (site-specific sump and/or tank) levels. 2. a. RPV water level unknown AND

b. Loss of RPV inventory per Table C3 indications.

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

NEI 99-01 Rev 6 Proposed EAL Justification HGl HGI flf f- ifeene {* Dvito Initiating Condition: GENERAL EMERGENCY Initiating Condition: iG No Change J Difference LJDeviation 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 Operating Mode Applicability: Operating Mode Applicability: timely classification.

All 1,2, 3, 4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift 1. A notification from the Security Force that a HOSTILE ACTION is occurring or has supervision). occurred within the PROTECTED AREA.

AND AND 2. a. ANY Table H1 safety function cannot be controlled or maintained.

b. EITHER of the following: OR

b. Damage to spent fuel has occurred or is IMMINENT

1. ANY of the following safety functions cannot be controlled or maintained.

Table H1 Safety Functions

• Reactivity control . Reactivity Control

• Core cooling [PWR] / RPV water level [BW4R] (ability to shut down the reactor and keep it shutdown)

RCS heat removal . RPV Water Level (ability to cool the core)

• RCS Heat Removal (ability to maintain heat sink)

OR

2. Damage to spent fuel has occurred or is IMMINENT Page 52 of 66

NEI 99-01 Rev 6 Proposed EAL I Justification HSI Initiating Condition: SITE AREA EMERGENCY Initiating Condition:

HSI M No Change 1111] Difference [-] Deviation HOSTILE ACTION within the Protected Area. HOSTILE ACTION within the Protected Area. 1) List site security shift supervision as Security Force.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

A HOSITLE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-security shift supervision). A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

Page 53 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: ALERT Initiating Condition: H No Change I Difference FIDeviation HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack 1) List site security shift supervision as Security Force.

threat within 30 minutes. threat within 30 minutes.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

1. A HOSTILE ACTION is occurring or has occurred within the OWNER 1 A validated notification from NRC of an aircraft attack threat < 30 minutes CONTROLLED AREA as reported by the (site-specific security shift from the site.

supervision).

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

2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.

Page 54 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HUl HUI Initiating Condition: UNUSUAL EVENT Initiating Condition: [F No Change F-- Difference L Deviation Confirmed SECURITY CONDITION or threat.

Confirmed SECURITY CONDITION or threat. 1) List site security shift supervision as Security Force.

Operating Mode Applicabllity: 2) Further described credible security threat through listing a site specific Operating Mode Applicability: procedure.

All 1, 2, 3,4,5, D Example Emergency Action Levels: (I or 2 or 3)

Emergency Action Level (EAL):

1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision). 1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.

2. Notification of a credible security threat directed at the site. OR

2. A validated notification from the NRC providing information of an aircraft

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

threat.

OR

3. Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.

Page 55 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HS6 HS2mmm Initiating Condition: SITE AREA EMERGENCY Initiating Condition: iS2 No Change I[ ] Difference [ Deviation Inability to control a key safety function from outside the Control Room. Inability to control a key safety function from outside the Control Room. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would Operating Mode Applicability: Operating Mode Applicability: enter should such an event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate All 1,2, 3, 4, 5, D shutdown were to be listed,

2) Added descriptors to better explain each safety function and allow for a Example Emergency Action Levels: (1 and 2) Emergency Action Level (EAL): timely classification.

Note: The Emergency Director should declare the Site Area Emergency Note:

promptly upon determining that (site-specific number of minutes) has 3) Changed "An event" to" A Control Room evacuation" to remove confusion been exceeded, or will likely be exceeded.

• The Emergency Director should declare the event promptly upon determining that if partial plant control was transferred to outside the control room with the the applicable time has been exceeded, or will likely be exceeded. control room still manned, due to testing or equipment failure.

1. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). 1 A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per

2. Control of ANY of the following key safety functions is not reestablished

• QCARP 0050-01, SB-i-1 Injection with SSMP and Bringing the Unit to within (site-specific number of minutes). Cold Shutdown OR Reactivity control . OCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to

• Core cooling [PWR] /RPV water level [BWR] Cold Shutdown

• RCS heat removal OR

. QOA 0010-05, Plant Operation with the Control Room Inaccessible AND

2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.

Table H1 Safety Functions

• Reactivity Control (ability to shut down the reactor and keep it shutdown)

• RPV Water Level (ability to cool the core)

• RCS Heat Removal (ability to maintain heat sink)

Page 56 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HAS H'A2 mmm Initiating Condition: ALERT Initiating Condition: [J No Change [j- Difference FIDeviation Control Room evacuation resulting in transfer of plant control to altemate locations. Control Room evacuation resulting in transfer of plant control to alternate locations. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would enter should such an Operating Mode Applicability: Operating Mode Applicability: event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate shutdown were to be listed, All 1, 2, 3,4,5, 0

2) Changed "An event" to" A Control Room evacuation" to remove confusion if partial plant control was transferred to outside the control room with the control Example 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 has resulted in plant control being transferred from the (site-specific remote shutdown panels and local control stations). A Control Control Room Room to evacuation alternate locations per:

• QCARP 0050-01, SB-i-i Injection with SSMP and Bringing the Unit to Cold Shutdown OR

• QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR

• QOA 0010-05, Plant Operation with the Control Room Inaccessible Page 57 of 66

NEI 99-Cl Rev 6 1 Proposed EAL Justification HU4 HU3 Initiating Condition: UNUSUAL EVENT FIRE potentially degrading the level of safety of the plant.

Initiating Condition: M No Change I Difference FIDeviation FIRE potentially degrading the level of safety of the plant. 1) Listed site specific list of plant rooms or areas that contain SAFETY SYSTEM equipment to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

All 1,2,3,4,5, D Example Emergency Action Levels: (1 or 2 or 3 or 4)

Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been Note:

exceeded, or will likely be exceeded.

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

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

• Report from the field (i.e., visual observation) A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of the

• Receipt of multiple (more than 1) fire alarms or indications following FIRE detection indications:

• Field verification of a single fire alarm

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

AND

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

b. The FIRE is located within ANY of the following plant rooms or

• Field verification of a single fire alarm areas: Table H2 Vital Areas (site-specific list of plant rooms or areas)

• Reactor Building (when inerted the Drywell is exempt)

• Main Control Room Envelope

2. a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).

• Unit and Shared Emergency Diesel Generator Rooms AND

• 4KV Switchgear Area

b. The FIRE is located within ANY of the following plant rooms or

• Battery Rooms areas:

• RHR Service Water Vaults (site-specific list of plant rooms or areas)

• Turbine Building Cable Tunnel AND

• Cribhouse

c. The existence of a FIRE is not verified within 30-minutes of alarm OR receipt.

2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of a FIRE).

3. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the AND plant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. b. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.

OR

3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-

4. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the minutes of the initial report, alarm or indication.

plant Protected Area] PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish. OR

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

Page 58 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT Initiating Condition: LJ No Change M Difference Deviation Seismic event greater than OBE levels. Seismic event greater than OBE levels. 1) Used Alternate developer notes allowed wording since specific Control Room indication of a seismic event > OBE is not available.

Operating Mode Applicability:

Operating Mode Applicability:

All 1,2,3,4, 5, D Example Emergency Action Levels:

Emergency Action Level (EAL):

Seismic event greater than Operating Basis Earthquake (OBE) as indicated by:

a. (site-specific indication that a seismic event met or exceeded OBE limits) 1. Control Room personnel feel an actual or potential seismic event.

AND

2. The occurrence of a seismic event is confirmed in a manner deemed appropriate by the Shift Manager or Emergency Director.

Page 59 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HA5i HA5if~f~

Initiating Condition: ALERT Initiating Condition: [AS M No Change [,, Difference [F, 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 to operations, cooldown or shutdown. cooldown or shutdown. ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example 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 is warranted.

• If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no emergency classification is

1. 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 H3 AND Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability

b. Entry into the room or area is prohibited or impeded.

Reactor Building* Mode 3 and 4 Turbine Building* Mode 3

-Areas required to establish shutdown cooling AND

2. Entry into the room or area is prohibited or impeded.

Page 60 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HU3 D]

HU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change n Difference -] Deviation Hazardous Event 1) Included River water level as part of the site specific list of natural or Hazardous Event technological hazard events. The EAL values selected are the current approved Operating Mode Applicability: EALtUE values.

Operating Mode Applicability:

All 2) Changed the word "needed" to "required by Technical Specifications" in the EAL 1,2, 3,4, 5, D to be consistent with terminology used by operators and minimize confusion.

Example Emergency Action Levels: (1 or 2 or 3 or 4)

Emergency Action Level (EAL):

Note: EAL #3 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Note:

I. A tornado strike within the PROTECTED AREA.

• EAL #4 does not apply to routine traffic impediments such as fog, snow.

ice, or vehicle breakdowns or accidents.

2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.

1. Tornado strike within the PROTECTED AREA.

3. Movement of personnel within the PROTECTED AREA is impeded due to OR an offsite event involving hazardous materials (e.g., an offsite chemical

2. Internal room or area flooding of a magnitude sufficient to require manual or spill or toxic gas release). automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode.

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

3. Movement of personnel within the PROTECTED AREA is impeded due to an

5. (Site-specific list of natural or technological hazard events) offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Abnormal River level, as indicated by EITHER:

a. High river level > 594 ft.

OR

b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam #14 has failed.

Page 61 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HG7 HG7 Initiating Condition: GENERAL EMERGENCY Initiating Condition: [i No Change U Difference U Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. declaration of a General Emergency.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT Other conditions exist which in the judgment of the Emergency Director indicate substantial core degradation or melting with potential for loss of containment that events are in progress or have occurred which involve actual or IMMINENT integrity or HOSTILE ACTION that results in an actual loss of physical control of substantial core degradation or melting with potential for loss of containment the facility. Releases can be reasonably expected to exceed EPA Protective Action integrity or HOSTILE ACTION that results in an actual loss of physical control of Guideline exposure levels off-site for more than the immediate site area. the 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

NEI 99-01 Rev 6 Proposed EAL Justification HS7 HS7 m D ne Initiating Condition: SITE AREA EMERGENCY Initiating Condition: L 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 warrant declaration of a Site Area Emergency. declaration of a Site Area Emergency.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example 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 indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that events are in progress or have occurred which involve actual or likely major that results in intentional damage or malicious acts; (1) toward site personnel or failures of plant functions needed for protection of the public or HOSTILE ACTION equipment 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 or access 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 not Guideline exposure levels beyond the site boundary. expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Page 63 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: ALERT Initiating Condition: [jj 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 warrant declaration of an Alert. declaration of an Alert.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2,3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential Other conditions exist which in the judgment of the Emergency Director indicate 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 that

~involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be involves probable life threatening risk to site personnel or damage to site limited to small fractions of the EPA Protective Action Guideline exposure levels, equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Page 64 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT Initiating Condition: LUh No Change [j Difference Deviation Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT. Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT.

Operating Mode Applicability:

Operating Mode Applicability:

All 1, 2, 3,4, 5, D Example 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 indicate degradation 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 facility response or monitoring are expected unless further degradation of safety systems protection has been initiated. No releases of radioactive material requiring offsite occurs. response or monitoring are expected unless further degradation of safety systems occurs.

Page 65 of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification E-HU1 E-HU1 Initiating Condition: UNUSUAL EVENT Initiating Condition: F1 No Change 1-- Difference FIDeviation Damage to a loaded cask CONFINEMENT BOUNDARY. 1) Listed 2x the site specific cask specific allowable radiation level as per Damage to a loaded cask CONFINEMENT BOUNDARY.

Certificate of Compliance No. 1014 Appendix A Section 5.7.

Operating Mode Applicability:

Operating Mode Applicability:

All 1, 2,3,4,5, D Example Emergency Action Levels:

Emergency Action Level (EAL):

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-technical specification allowable radiation level) on the surface of the spent fuel contact radiation reading:

cask.

• > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR

• > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts.

Page 66 of 66

I=xAInn N.nlmar Oijad Citiesa Annex Exelnn NucleIuar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARGI Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRfem TEDE or 5000 mRirem thyroid CDE.

Operating Mode Applicability:

1, 2,3,4,5, D Emergency Action Level (EAL):

Notes:

" The Emergency Director should declare the General _mer=.gency 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 path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

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

(1) Reading on ANY of the following radiation monitor-s greater than the reading shown for 15 miueso longer:ý (Site Specific monitor list and threshold v:alues)

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCilsec for >

15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR

2. Dose assessment using actual meteorology indicates doses at or beyond (site-speGifi dose receptor POW.) the site boundary of EITHER:

a. > 1000 mRem TEDE OR

b. > 5000 mRem CDE Thyroid OR Month 20XX QC 3-1 EP-AA-1006 (Revision XX)

r1iiesA fifimom Annmv IP=alnn Mnilanr S.'U;I l *ll rl=III 5fl kA~l.I 11 .51 l il RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS FWild- SU y...results indicate EITHER of the following at OF beyond (site specific dose

,e veptOFv, . ,, -.

Closed window do6e rates greater than 1,000 mR'~hr expected to continue for 60 minutes or longer.

Analyses of field- survey sampler, indicato thyroid CDE greater than 5,000 mrFem:

3. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates >1000 mR/hr are expected to continue for > 60 minutes.

OR

b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

The TEDE dose is set at the EPA PAG of 1000 mRrem while the 5000 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor rcadings assumes that. a release path to the enlVironment is established. if the effluent flow past an effluent monGitor is known to hav stopped due to actions to isolate the release path, then the effluent mon0itor reading is-no longer Valid. f9or clasAsi-ation purposes.

Basis Reference(s):

1. NEI 99-01 Rev 6, AG1

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-2 EP-AA-1006 (Revision XX)

(I"orl Mfiaa Annov =vm1en kimie-Igftr flu~~~~~_ý.v Fý"rffI.m.I~ vvIIiio~wz RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARS1 Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mRfem TEDE or 500 mRfem thyroid CDE.

Operating Mode Applicability:

1, 2,3, 4, 5, D Emergency Action Level (EAL):

Notes:

" The Emergency Director should declare the Site Area Emergency 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 path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

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

(1) Reading on ANY of the WoloG.ing radiation monitors greater than the reading shown fQr 15 minutes Or Ionger:

(Site spc oii aInd threshold values)

(2) Dorse assessment using actual meteorology indicates doses greater than 100 mr*em TEDE or 500 mr*eFm thyrod DE at or beyond (,ite Sp;cifc dose r*e*eptr (3) Field survey reaults indicate EITHER of the fllimneg at Or beyond (>ite EpeCific dose receptor poGint):

• Closed window dose rates greater than 100 mR'hrf expeeted to coentinlue for-60 minutes er-lengefn 9, Analyses of field survey samples indicate thyroeid CDE greater-than 500 mrwem for-one hour-of inthalation.

3. 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCllsec for > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR Month 20XX QC 3-3 EP-AA-1006 (Revision XX)

flimnrl f'ifidsa Annghv I::vy--Inn k,,rl--ghr nil an (~i*rne Annav Fva Ii~n him u,.Iaar RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

a. > 100 mRem TEDE OR

b. > 500 mRem CDE Thyroid OR

3. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates >100 mR/hr are expected to continue for > 60 minutes.

OR

b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite 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 are associated 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 and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1000 mRrem while the 500 mRfem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classific~ation based on effluent monitor readings assumes that a release path to the enVIronment is established.If the effluent filw past aI effluent onitoris known to have stopped due to actions to isolate the ree Ipah, then the effluent monAitor reading is 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, AS1 Month 20XX QC 3-4 EP-AA-1006 (Revision XX)

Exalon Nuclear Quad Cities Annex Eeo ula RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-5 EP-AA-1006 (Revision XX)

t'limarl Pifiae Annnv I=:y--nn N~t-lazar Amman r~it;~ Ann~w Fv~Inn I~J..vIoar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA1 I Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mR-em TEDE or 50 mRfem thyroid CDE.

Operating Mode Applicability:

1, 2, 3, 4, 5, D Emergency Action Level (EAL):

Notes:

• The Emergency Director should declare the Ale4-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 path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

• The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. rll for 15 minutes Or longer:.

(site sperific monitor list and threshold value&)

(2) Do.e asserssment using ac.tual meteero1rgY i.ndiates doses. greater thaRn 10 mRm..

TEDE or 50o mrem thyro-id C-DEM at or bevond (site se~ecfi dose Fece~tGr Do9intL I l *

(3) Analysis of a liquid effluent sample indieates a conc-entrantion or releas14e rate that would result in doses greater than 10 FW.mrvm TEDE Or 50 mne. thyroid MDE At or beyond (Site specific doerc Ptooint) for one hour of texposure.

(4) Field survey results indicateEITHER of the following at or beyond (cite.pecific doserecepto pint):.

" Closed window dose rats gr-eater-than 10 mR'hrexpected to continue for-60 minutes or

" Analyses of field sur.'ey samples iadiceAc thyroid CDE gr-eater thin 50 ffwem for-oef hour Of inhalatien.

Month 20XX QC 3-6 EP-AA-1006 (Revision XX)

Exalon Nuclaar Ouad Cities Annex Eeo ula RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 uCilsec for > 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

a. > 10 mRem TEDE OR

b. > 50 mRem CDE Thyroid OR

3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary

a. 10 mRem TEDE for 60 minutes of exposure OR

b. 50 mRem CDE Thyroid for 60 minutes of exposure OR

4. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for_> 60 minutes.

OR

b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual 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 this magnitude represent an actual or potential substantial degradation of the level of safety of Month 20XX QC 3-7 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS the 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 and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more 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 mRfem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classific-ation based on offluont monitor readings assumes, that a roloase path t the *1nim e*1Int is establihe. If the AMlu, nt flow part an effluent monit is~l known to have  ;*~~.

... due stopped ';. to- ac--9tion to, igsolate-

.-rc

.. the

-.. release path, -.

then the effluen~t mon~itOF read~ing4",."9*"**"p is noe longer valid for cliiaIo proe.S Escalation of the emergency classification level would be via IC RAS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA1

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-8 EP-AA-1006 (Revision XX)

Ouad Cities Annex Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARUM Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM (site-speifir effluent releasecontr*olling document) limits for 60 minutes or longer.

Operating Mode Applicability:

1, 2, 3, 4, 5, D Emergency Action Level (EAL):

Notes:

" The Emergency Director should declare the Unusual Event event promptly upon determining that 60-iiwtesthe 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 60 minutes.

" Classification based on effluent monitor readings assumes that a release path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

(1) Reading on ANY effluenRt radiation monitor greater than 2 times the (site specific effuen reease cOntrolling documnent) limits for 60 mninutes or longer:-

(site specific-monitor li6t and threshold values corresponding to 2 times the controlling docum~ent limits)

(2) ReadinRg on ANY e-ffluenAt radiation monFitor groAator- than 2times the alarm setpin et-ablished by a current radioactivi,' discharge permit fogr 6-0) minuItes6 Or longer.

(3) Samnple analysis for a g~asouso;r liquid release indicates a conentratieono repleamse ra-te greater than 2 times, the (site spec-ific effluenA-t release controlling document) limits-for-6-0 mninutes Or longer.

1. Reading on any of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for 2!60 minutes.

• Radwaste Effluent Monitor 1/2-1799-01 OR

" Discharge Permit specified monitor OR

2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined Control Room Panels or PPDS - Total Noble Gas Release Rate).

Month 20XX QC 3-9 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS LVACIand SGrIS Rn4dation onitor;r.1706 17 )2fond --- on Control room8 Panels-OF-OR

3. 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 a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological 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 of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

Classifiati-n based oR *oni*tGr effluent readi;ngs asumes that a Frlaso path the enyiro~nmePnt_ is established-. If the effluent- flowt pact an efflueAnt_ moitri kno~Wn to hav tpped due to actions to isolate the release path, then the effluent mon;itor reading is no longer valid foar c~lassific-atio pupoe.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

EAL #1 Basis EAL--#2-- This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.

EAL #2 Basis EAL-#-I-This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous OFrlilquid-effluent pathways.

Month 20XX QC 3-10 EP-AA-1006 (Revision XX)

(Iiind (Nfigm. Annoy l::xelnn N.PIA*r flmmnI Ci~cu Anv Fe~nn Nuclehar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS EAL #3 Basis

• AL- #3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC RAA1.

Basis Reference(s):

1. NEI 99-01 Rev 6, AU1

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-11 EP-AA-1006 (Revision XX)

I=xelon Nuclear Ouiad Cities Annex Exellon Nucleanr RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA2 I Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel.

Operating Mode Applicability:

1,2,3, 4,5, D Emergency Action Level (EAL):

=="=! 1K1Q- r3A-rU1A1AN/

'-I1 I 7 f t* *l .Iy* *I * * . . .

(2) Damage to irraidiaEtedC- fu-el reSUlting in a release of radioactivity from the fuel as inicte by 'ANY of the folloin~g radiation monitors:

(site specific liStinig Of radiation monitors, and the associated readings&,

setpoints and/40r allarm)

(3) Lowering of spent fuel pool level to (site specific Level 2 value). [320e Deve~opo Noto&J

1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

OR

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R1 Radiation Monitor reading >1000 mRemlhr Table R1 Fuel Handling Incident Radiation Monitors 0 1(2) 1705-16A Fuel Pool Rad Monitor 0 1(2) 1705-16B Fuel Pool Rad Monitor Basis:

REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.

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

CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.

Month 20XX QC 3-12 EP-AA-1006 (Revision XX)

OsiuI Cities Annex ExeIon Nuclanr RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly fuel

~ . ~a.. *"*"

~ orasnFiatýMW~

• oong assmby. ....-

. We " "--Of.-.....

Water ' .... within the levrel '-f** . spent

. . fel'" (seo poo!

D..veoe. N..tes). These events present radiological safety challenges to plant personnel 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 the plant.

This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.

Escalation of the emergency would be based on either Recognition Category A or C ICs.

EAL #1 Basis RAI *4 This EAL escalates from RAU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

EAL #2 Basis EAL-#2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, 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 a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency would be based on either Recognition Category RA-or C ICs.

Month 20XX QC 3-13 EP-AA-1006 (Revision XX)

O.iad Riesw Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Spent fuel pool Wator level at this value is Within the loWer And of the level range necessary to preVent significant dose consequencaes from direct gamrma radiation to personnel po~ferming Operations in the vicainity of the spent fuel pool. Thiscodtn refloc~te a significant losr, of spent fuel pool Water inventor; and thus iot. is aloa precursor to a loss of the ability to adequately cOOl the irradiated fuel assembles stored in the pol Escralat-ion of the Wemegency c-lassific~ation level would be via IQ; ASI1 or A82 (sag ASL2 Devoloe, ,oth).;,

Basis Reference(s):

1. NEI 99-01 Rev 6, AA2 Month 20XX QC 3-14 EP-AA-1006 (Revision XX)

n"nrl r-ifina Annav I=Y,-nn N, ielan-r

0. .~ ri (~i*igi@ Annoy Fvolnn MivIa~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARU2 Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.

Operating Mode Applicability:

1,2,3,4,5, D Emergency Action Level (EAL):

(1) a. UNPLAlNNED water level drop in the REFUELING PATHWAIAY as inrdicat by ANY of the following:

(site specific level indieatiens).

b. UNPLANNED rie inae adiation levels as indlicated by ANY of the following radiation monitors.

(site spocific list of area radiation menitor-s)

1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

" Refueling Cavity water level < 282 in. (Upper Wide Range simulated signal).

OR

• Spent Fuel Pool water level < 19 ft. above the fuel (<4 ft indicated level).

OR

• Indication or report of a drop in water level in the REFUELING PATHWAY.

AND

b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R1.

Table RI Fuel Handling Incident Radiation Monitors

• 1(2) 1705-16A Fuel Pool Rad Mon

• 1(2) 1705-16B Fuel Pool Rad Mon Month 20XX QC 3-15 EP-AA-1006 (Revision XX)

flivmA r'i#iga Anng%,w P:valnn Kii-rloor A. .~ ,I 1~.*rne Annn~v Fvair~n Mm.rIc~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.

This IC addresses a de*ease loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level deerease-loss will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available) or from any other temporarily installed monitoring instrumentation. A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance 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, AU2

2. Technical Specification 3.7.8

3. Technical Specification 3.9.6 Month 20XX QC 3-16 EP-AA-1006 (Revision XX)

flvaýA ('Mgs4m Anngav I::valnn ki~rl-,ldr RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA3 I Initiating Condition:

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

Operating Mode Applicability:

1, 2, 3, 4, 5, D Emergency Action Level (EAL):

Note:

Ifthe equipment in the listed-room or area listed in Table R3 was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted to Contr@Ioleefom G enftral Al---m Station 0 (other Site Specifi ar-eaSrooms) levels that I^\ A I I i I ll*l A ill Ip[l*

• JL--

6A)in AR WNPUNNNI-w event rosults 1n Faalialion WGRoENio Or iMDPeA RGracces t any; of the folloWing plant Fooms OF areas:

(site 6necific !irt of ...... F ilant reoms Or

v. areas with entr related Mode anpnmI*nallit

........... j V-I o

1. Dose rate geate--than> 15 mR/hr in ANY of the areas contained in Table R2:

Table R2 Areas Requiring Continuous Occupancy

" Main Control Room (Unit 1 ARM Station #22)

" Central Alarm Station - (by survey)

OR

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to any of the areas contained in Table R3:

Month 20XX QC 3-17 EP-AA-1006 (Revision XX)

flomri fifimcz Anngwv I:v,-Inn Mlill-,ar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Table R3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building* Mode 3 and 4 Turbine Building* Mode 3

• Areas required to establish shutdown cooling Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 to preclude or impede personnel from performing actions necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal plant proceduresmaintain norm.al plant operation, Or to perform a normF.al plant c .oldownan sh'tdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

Table R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown. This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.

For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in serviccat the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding beyond that required by procedures, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

Month 20XX QC 3-18 EP-AA-1006 (Revision XX)

Exalon Nuclear ExlnNcEar Ouad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS 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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

" The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

" The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

" The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

Escalation of the emergency classification level would be via Recognition Category RA, C or F ICs.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA3

2. FSAR Section 3.2 Month 20XX QC 3-19 EP-AA-1006 (Revision XX)

nmvmA ifigsim Annghw i::Ylnn N,,P-lan::r Am.~rI f~4a~ Aatui~w Fvair~n Mi urIn~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS SRU3 Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

(I) (Site specific. radiation moenitor) reading greater than (site specific value)-.

(2) Sample analysis indicates that a reactorcoolant actvity value as greater than an a.llow-able Im;it in Technical Spcfi÷atiOnc.

pe,ified

1. Offgas system radiation monitor HI-HI alarm.

OR

2. Specific coolant activity > 4.0 uClIgm Dose equivalent 1-131.

Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified 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 clad degradation should not result in the declaration of an Unusual Event.

This EAL addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity.

An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample activity (as determined by laboratory confirmation). Fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another 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, SU3

2. Technical Specifications 3.4.6

3. Technical Specifications 3.7.6 Month 20XX QC 3-20 EP-AA-1006 (Revision XX)

A. .21 (difiaa Annav I~vnlt~n k~l-I-mr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:

Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.

Operating Mode Applicability:

1,2,3 Emergency 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 I4Table 9-F-2 Month 20XX QC 3-21 EP-AA-1006 (Revision XX)

(1"nrl ('ifiaa Annav I:va,-nn K,,r-l,--r A. .~ ri t~itic~@ Annoy Fvolnn N. .rIa~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:

Loss or Potential Loss of ANY two barriers.

Operating Mode Applicability:

1,2,3 Emergency 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-3Table 9-F-2 Month 20XX QC 3-22 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FAI Initiating Condition:

ANY Loss or ANY Potential Loss of either Fuel Clad or RCS.

Operating Mode Applicability:

1,2,3 Emergency 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 more heavily 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 of Containment barrier in combination with loss or potential loss of either Fuel Cladding or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F 3Table 9-F-2 Month 20XX QC 3-23 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:

RCS Activity Operating Mode Applicability:

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

LOSS A. (Site Spcciftic ndicatiens that reactor coolant activity is grcatcr than 300uGi/gmR doc

...... *quivI 3!)Coolant activity > 300 uCilgm Dose Equivalent 1-131.

Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 ýiCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-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-9F-4Table 9-F-2 Month 20XX QC 3-24 EP-AA-1006 (Revision XX)

l=x*lon NuclAar Oijad Cities Anne~x Exelnn Nuclea~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS A-.-1. Plant conditions indicate Primary Crontainment flooding is required.

POTENTIAL LOSS A-.2. RPV water level cannot be restored and maintained abovo (site specific RPV wateF. level GOg

...... .. hE , . fuel)..,,,> -142 inches (TAF)

3. RPV water level cannot be determined.

Basis:

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

Loss 2AThreshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.

This is identified in the BWROG F-PGsEOPs/SAGsSAMGs when the phrase, "Primary Containment Flooding Is Required," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "Primary containment flooding required,"

also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.

Potential Loss 2-,AThreshold #2 and #3 Basis This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.

The RPV water level threshold is the same as RCS baF-ei-Barrier Loss threshold-24A.

Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier 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, RPV-wat&RPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the Month 20XX QC 3-25 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order 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 drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

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

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 to deliberately lower RPV water level to the top of activ" fuel in order to reduce reactor power. RPV water level is then contro.lled beteen the top of activ; fu!l and the Minimum Steam Cooling RPV V^ater Level (..S.RV*ý.- Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA5-MA3 or SS5-MS3 will dictate the need for emergency classification.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, Table9 F4,Table 9-F-2 Month 20XX QC 3-26 EP-AA-1006 (Revision XX)

MwnA Pifigsim Annmv I:vylnn kielaar Am. a ~4 r~;4i~a Anr~aw ~vciInn MmmI~ar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS A. Primary containment radiatonR MGnitor reading greater than (site specific. value)

Drywell radiation monitor reading > 6.65 E+02 R/hr.

Basis:

L-ess-4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 [tCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of 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 RCS Barrier RC5 Loss Tthreshold 44-Asince it indicates a loss of both the Fuel Clad Barrier and 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 Potential Loss threshold associated with Primary Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-9-F-4Table 9-F-2

2. Core Damage Assessment Methodology Month 20XX QC 3-27 EP-AA-1006 (Revision XX)

Exalon Nuclear Ouad Cities Annex Exe.... Nu.. e..

RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS 1A. Any condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.

POTENTIAL LOSS 2A. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.

Basis:

Loss Threshold #1 Basis L*,*&A This 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 #2 Basis Petentwal Io L.,6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table -F-3Table 9-F-2 Month 20XX QC 3-28 EP-AA-1006 (Revision XX)

Mind (Nfiaa Annoy IFyalnn khielanr flu .~ rI ('~i*i~@ Ann~w Fvalnn hi.., ia2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS

1. RPV water level cannot be restored and maintained abovo (site "pecific RPV water level correpondin't the top of active fuel)

'> -142 inches (TAF)

Gr-OR

2. RPV water level cannot be determined.

Basis:

Lgs&-2L.A RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.

The RPV water level threshold is the same as Fuel Clad bairieF Barrier FC2 Potential Loss threshold--2-A. Thus, this threshold indicates a Loss of the RCS barrier and Potential 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, RPV-wateFRPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order 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 drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The Month 20XX QC 3-29 EP-AA-1006 (Revision XX)

Exelan NUC102r Oijnd Cities Annex Exl... Nu.le.r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

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 to deliberately lower RPV water level to the top of arct'e fue-l in order to reduce reactor power. RPV water leVel i6 then controlled between the top of a.t.ve fuel and the Minimum Steam.Co

• oling RPV Water Le'vel (MSCGRV^AW.Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA5-MA3 or SS&MS3 will dictate 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-§-F-3Table 9-F-2 Month 20XX QC 3-30 EP-AA-1006 (Revision XX)

Clmiorl fifiaa Anng%,v IPv--Ign khme-l-or A. marl r~;*;~ Anv~av Fv~Lnn Mm.iIaar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

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

LOSS A. Primary containment pressure greater than (site specific"value) due to RCS leakage-.

1. Drywell pressure >2.5 psig.

AND

2. Drywell pressure rise is due to RCS leakage Basis:

The (site spe.ific.value)> 2.5 psig primary containment pressure is the 4ryweI4Drywell high pressure setpoint which indicates a LOCA by automatically initiating the-ECCS-of oguivalent makeup system.

The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.

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

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications 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---F--3Table 9-F-2 Month 20XX QC 3-31 EP-AA-1006 (Revision XX)

rlimexri ('Mga w~s~ ~

Annimv

,-~. .5 .~n Fval~nn

~

kuir ibr S S RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:

RCS Leak Rate Operating Mode Applicability:

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

LOSS Al. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC line break. in ANY of the following: (site specific. systems with potential for high energy line OR

82. Emergency RPV Depressurization is required.

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

a4. Secondary Containment area temperature > QGA 300 MaximumMax Normal Ooperating Tempe~akelevels.

OR b-2. Secondary Containment area radiation level > QGA 300 Maximum Max-Normal Ooperating Area Radiation ,Level.

Basis:

UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.

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 that the on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.

Month 20XX QC 3-32 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Loss Threshold #1 Basis-3A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the CoGntrol Room, the RCS barrier Loss threshold is met.

Loss Threshold #2 Basis-4B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the 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 the RCS is being vented into the Torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

Potential Loss Threshold- #3 Basis 3-1A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.

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

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam 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 to a Site Area Emergency when combined with Containment Barrier CT6 Loss Tthreshold

1. 13A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Basis Reference(s):

Month 20XX QC 3-33 EP-AA-1006 (Revision XX)

Ouad Citie Annex Ex.lon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION

1. NEI 99-01 Rev 6, 1b-9-FTable 9-F-2 Month 20XX QC 3-34 EP-AA-1006 (Revision XX)

flmitsri ('ifina Annnv i=Y'*h*n kJi iijhmr Au. ~ ii t~i*iao** Anna~v r~. ** 3s 1S fl

• RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:

Primary Containment radiation Operating Mode Applicability:

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

LOSS

1. Drywell radiation monitor reading > I00R/hr.

A. Primary containment radiation reading greatcr than (Site pecific value).

Basis:

Less-4-A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 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 RCS Barrier only.

There is no Potential Loss threshold associated with Primary Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-9-F-3Table 9-F-2

2. Calc. EP-EAL-0611 Month 20XX QC 3-35 EP-AA-1006 (Revision XX)

tlmmA Mfiaa Annav I::valr'*n Klir-lanr flu a~ rI (~j*jgai@

~t4a . .. Ann~v rn .. .n Fvalnn S**U *Miur6~r

• • ~c~

RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS Al. Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.

POTENTIAL LOSS A2. Any condition in the opinion of the Emergency Director that indicates Potential I Loss of the RCS Barrier.

Basis:

Loss 6-AThreshold #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 #2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-Q-F4Table 9-F-2 Month 20XX QC 3-36 EP-AA-1006 (Revision XX)

Onnrl Cities Annexw ExeIon Nucleanr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

POTENTIAL LOSS A.-Primary containment flooding is required.

Basis:

Potential Loess 2.A.

The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGsEOPs/SAGsSAMGs specify the conditions that require primary containment flooding. When primary containment flooding is required, the EPGsEOPs are exited and SAGsSAMGs are entered. Entry into SAGsSAMGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt 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 a General Emergency.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-9-F4Table 9-F-2 Month 20XX QC 3-37 EP-AA-1006 (Revision XX)

Ouad Cities Annex Exelon Nuclear I=x*lon Nucilear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Primary Containment Conditions Operating Mode Applicability:

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

LOSS Al. UNPLANNED rapid drop in prmry .tainmentDrywell pressure following P.a GentaimentDrywell pressure rise.

OR B2. Prim.a.y contai.me nDrywell pressure response not consistent with LOCAI conditions.

POTENTIAL LOSS A3. Prim.ay contai÷nmntDrywell pressure greater than (site specific Value)> 56 psig and rising.

OR

84. (site s eXPlosive mixture) exists inside prima.y containment a. Drywell or

.peci..

torus hydrogen concentration > 6%.

AND

b. Drywell or torus oxygen concentration > 5%.

OR G5. HTLG-Heat Capacity Limit (QGA 200, Figure M)exceeded.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Loss !-A-a n-d-,BThreshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to dPywellDrywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity. Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. 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 the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment Month 20XX QC 3-38 EP-AA-1006 (Revision XX)

f-l"nel Mae Annoy Pgi:n~rn Nh:PlAd-ar A. .~ .1 (~i$i~o Annoy Fvolnn ItJ,,rIo~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.

Potential Loss 4-tAThreshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.

Potential Loss 4.BThreshold #4 Basis If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.

Potential Loss 4-.GThreshold #5 Basis The HCat Capacity Temperatufe Limit (HpTr) is the highest Torus twmperature from Eergency RP2V D)Pep-ressu-ri-atinn will no-t raais-e:-

which equipmcnt within the Torus which may be required to operate whenM t~he RPV is-

• Torus pressure aboeve Priimar,' Containment PreSSUre Limit A, whilec the rate of energy transfer from the RP2V to the containment is greater than the capacityo the GentainmenRt vent.

The HCTL is a function of RPV pressure, Torus temperature and 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 and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

Basis Reference(s):

1. NEI 99-0 1 Rev 6, Table 9 I=3Table 9-F7-2 Month 20XX QC 3-39 EP-AA-1006 (Revision XX)

r1minel Pifigaa Annnv IPv,-Inn M, iP-lar Am, ~ rI(*iniaAnnn~v FvaIr~n 1Ji.rIa~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

POTENTIAL LOSS A.PriMaR' containmel# radiation monGitor reading greatker than (site Spocvific value)

1. Drywell radiation monitor reading > 1.55 E+03 R/hr.

Basis:

There is no Loss threshold associated with Primary Containment Radiation.

Poten~tial Loss 4. A.

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. 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 NuclearPower Plant 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 exist- there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential 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-3Table 9-F-2

2. Core Damage Assessment Methodology Month 20XX QC 3-40 EP-AA-1006 (Revision XX)

flimrl ifima Annmv IB'vg=nn Ki, lorl'r Al. u

~ *i*ia@

t* Annaw St**U fl Fvc S** innS UNh.rIc~r

.1, RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:

Primary Containment Isolation Failure Operating Mode Applicability:

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

LOSS Al. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR B2. Intentional Pprimary CGontainment venting/purging per EOP's or SAMGs due to accident conditions.

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

1. Secondary Containment area temperature > QGA 300, MaximumMa* Safe Qoperating TenpeFakwelevels.

OR

2. Secondary Containment area radiation level > QGA 300, MaximumMa* Safe Ooperating Radiatio* I-levels.

Basis:

UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.

These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss 3-.AThreshold #1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument 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 breached and thereby creates a significant pathway to the environment. Examples include unisolable Main Steamline, HPCI or RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.

Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Month 20XX QC 3-41 EP-AA-1006 (Revision XX)

Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. 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 but should be evaluated using the Recognition Category A-R ICs.

Loss 3,BThreshold #2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the dPywelDrywell high pressure scram setpoint) does not meet the threshold condition.

Loss 3.CThreshold #3 Basis The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each 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 the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam 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 Lioss &AThreshold #3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Basis Reference(s):

Month 20XX QC 3-42 EP-AA-1006 (Revision XX)

C. utsv4 Mfid~a Annmv~

-~ 5. i=yalnn N"Aal*nr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION 1 NEI 99-01 Rev 6, Table-Q-FTable 9-F-2 Month 20XX QC 3-43 EP-AA-1006 (Revision XX)

flimri Pifigha Annghv l~v--lnn klmrI-,ar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS Al. Any condition in the opinion of the Emergency Director that indicates Loss of the I Containment Barrier.

POTENTIAL LOSS A2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.

Basis:

Loss $-AThreshold #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 #2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table9 F- 3Table 9-F-2 Month 20XX QC 3-44 EP-AA-1006 (Revision XX)

rlimýri ('Mga Anngv Fwi:vlnn NllrAaar Au ~rI r~;*;~ Ann~v Fv~Inn Mmmr.6~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSG1 Initiating Condition:

Prolonged loss of all Off-site and all On-Site AC power to emergency busses.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emerge..yevent promptly upon determining that (site specific hOurS) the applicable time has been exceeded, or will likely be exceeded.

1---a. Loss of ALL offsite and- ALL, on'it- AC power to ("it Specific em.ergen.y buses)unit ECCS busses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND 3b. EITHER of the following:

a. Restoration of at least one em&§enGy-unit ECCS bus in < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is notles.

than speGift, heus-i

.(site . 'net likely.

OR

b. RPV water level cannot be restored and maintained > -166 inches.

(Site specific indic-ation o-fan inability to adequately remRove heat fromn the core)

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat Month 20XX QC 3-45 EP-AA-1006 (Revision XX)

rl"nrl (ifiaa Annov P~valnn khnnllanr flu .~ r4 (~i*iao Ann~v Fv~Inn Mumr1~2r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of -- o Or mo -- ny fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event 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 not be used as a basis for delaying a classification upgrade. The goal is to maximize the time 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 power results in parameters that indicate an inability to adequately remove decay heat from the core.

Basis Reference(s):

1. NEI 99-01 Rev 6, SG1

2. UFSAR 8.3 Month 20XX QC 3-46 EP-AA-1006 (Revision XX)

Dinid Citiest Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY M$SI Initiating Condition:

Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Sito Area r*,-..genc-yevent promptly upon determining that the applicable time 165- inues-has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite and ALL -n'itoA*C Power to (site specific emnergeny bus.s)unit ECCS busses fQo 15 minutesorO lonr.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RAG1, FG1 or MSGI.

Month 20XX QC 3-47 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis Reference(s):

1. NEI 99-01 Rev 6, SS1

2. UFSAR 8.3 Month 20XX QC 3-48 EP-AA-1006 (Revision XX)

IExAlnn Nuclanr 0"nd Cifian Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSA1 Initiating Condition:

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

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the eventAled promptly upon determining that the applicable time 4§5rFinutes-has been exceeded, or will likely be exceeded.

1. AC power capability to unit ECCS busses reduced to only one of the following power sources for > 15 minutes.

• Reserve auxiliary Transformer TR-12 (TR-22)

• Unit Auxiliary Transformer TR-1 1 (TR-21)

• Unit Emergency Diesel Generator

• Shared Emergency Diesel Generator

• Station Blackout Diesel Generator

• Unit crosstie breakers

a. AC PoWer capability to (site specific e*ne*geRnY buses) is re-dued, to a single power Source forF 15 minutes or longer.

AND 2b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC MSU1.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

Month 20XX QC 3-49 EP-AA-1006 (Revision XX)

Mind Civinv An.n. Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY

" A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

" A loss of all offeite power and- lo9s86 Of a~ll emer9gency power soreFes (o.g., GAit die-Sel generators)-' wih snle train of emergencay buses, being back fed from th Unit mai gnrtor.

" A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Escalation of the emergency classification level would be via IC MSS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA1

2. UFSAR 8.3 Month 20XX QC 3-50 EP-AA-1006 (Revision XX)

ula Exelon Nuclear Eeo Ouad Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Msul Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unu'ual E,-cntevent promptly upon determining that the applicable time 4Min#ute&s-has been exceeded, or will likely be exceeded.

-.Loss of ALL offsite AC power capability to unit ECCS busses (site-speGifiG

.....gen.y bue)

.. 9F for > 15 minutes-er- i.

Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders 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 power source(s) is available to the emergency buses, whether or not the buses are powered from it. (e.g. unit cross-tie breakers)

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC MSAI.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU1

2. UFSAR 8.3.1 Month 20XX QC 3-51 EP-AA-1006 (Revision XX)

Explon Nuclear Ouad Cities Annex Ex......uc...r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSG21 Initiating Condition:

Loss of all AC and Vital DC power sources for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the General *,e*,rgenr. vent promptly upon determining that the applicable time 4-5miRutes-has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to unit ECCS busses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to vital busses.

AND

3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.

AND

4. All AC and Vital DC power sources have been lost for >15 minutes.

buses) for 1 -5 minutesor lOFnger.

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• l*V J nnl A'*hhI yl I K (cite munific Vitl DvC hicerqw fio 15 min,udte or long.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

Month 20XX QC 3-52 EP-AA-1006 (Revision XX)

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%fA -aI. naw FvlnM.rI RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both-all EALs threshelds-are met.

Basis Reference(s):

1. NEI 99-01 Rev 6, SG8

2. UFSAR 8.3.2

3. Technical Specifications B.3.8.4 Month 20XX QC 3-53 EP-AA-1006 (Revision XX)

fluisd Cities Annex Exelon Nuclea~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month 20XX QC 3-54 EP-AA-1006 (Revision XX)

fliinrl fMae Anngi-tr =vgmlrn Klimr-lAmor Al I ~tI i4la~ Aijnav ~vaIr~n MumIn~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSS28 Initiating Condition:

Loss of all vital DC power for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Site Area Emergency-vent promptly upon determining that the applicable time 41-minutes-has been exceeded, or will likely be exceeded.

kIdi*ated-vVoltage is < 105 VDC less than (site specific. bus voltage value) on 125 VDC battery busses #1 and #2ALL (site specific V--ital D, busses) for >15 minutes OF.4Gnge

'Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a loss of Vital DC power which compromises the ability to monitor and 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 power losses.

Escalation of the emergency classification level would be via ICs RAG1, FG1 or MSG28.

'Basis Reference(s):

1. NEI 99-01 Rev 6, SS8

2. UFSAR 8.3.2.1.1

3. Technical Specifications B3.8.4 Month 20XX QC 3-55 EP-AA-1006 (Revision XX)

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%ftý RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSS35 Initiating Condition:

Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

2. All manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.

AND

3. EITHER of the following conditions exist:

• RPV water level cannot be restored and maintained > -166 inches OR

" Heat Capacity Limit (QGA 200, Figure M) exceeded.

(Site spec-ific- ind-ic-ation of an inability to adequately remov.e heat fromn the cree)

(Site 6pecifi indication of an inability to adequately remo-ve he-at frm.-. t~he RC;S)

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection,actins, to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

Month 20XX QC 3-56 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

Escalation of the emergency classification level would be via IC RAG1 or FG1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SS5 Month 20XX QC 3-57 EP-AA-1006 (Revision XX)

ExAIon Nuclear Ousari ities Annex Exelon Nucla~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSA36 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

1. An-aAutomatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

2. Manual / ARI actions taken at the reactr control cnc.s.clReactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged Month 20XX QC 3-58 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MSS35. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC MSS35 or FS1, an Alert 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 are included 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, SA5 Month 20XX QC 3-59 EP-AA-1006 (Revision XX)

O.,ad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSU31 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

1. a. Ap,-aAutomatic scram -did not shutdown the reactor as indicated by Reactor Power > 5%.

.AND

b. A-sSubsequent manual action taken at the reactor control consl Reactor Console is successful in shutting down the reactor.

OR

2. a. A-mManual scram / ARI did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

b. EITHER of the following:

1. A-sSubsequent manual / ARI action taken at the reactor control GeRseleeReactor Console is successful in shutting down the reactor.

OR

2. A-sSubsequent automatic scram / ARI -is successful in shutting down the reactor.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in 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 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

Month 20XX QC 3-60 EP-AA-1006 (Revision XX)

I=xelon Nuclear Q~mzrl iCigti Annex Exellon Nucleanr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor tscram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MSA35. Depending upon the plant response, escalation is also possible via IC FAI. 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 setpoint testing), the following classification guidance should be applied.

" If the signal generated as a result of plant work causes a plant transient that created a real condition that should have included an automatic reactor scram and the RPS fails 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 but should 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 EALs are not applicable and no classification is warranted.

Month 20XX QC 3-61 EP-AA-1006 (Revision XX)

Exelon Nuclear Quad Cities Annex Eeo ula RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis Reference(s):

1. NEI 99-01 Rev 6, SU5 Month 20XX QC 3-62 EP-AA-1006 (Revision XX)

QU2d Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSA42 Initiating Condition:

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

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the eventAle~ promptly upon determining that the applicable time 45 minutes-has been exceeded, or will likely be exceeded.

1. a. AR,-UNPLANNED event results in the inability to monitor one Or mRcANY Table Mlef the*f*llowing parameters from within the Control Room for > 15 minutes oF loRge-.

[see table below]

JB WR parameter&41t Reactor PoWer Table M1 Control Room Parameters

• Reactor Power R"V Water-Level

• RPV Water Level

• RPV Pressure "VX PrcSSUrc

• Drywell Pressure Primary Containment

• Torus Level

• Torus Temperature Suppression Pool Level SUPPrcssion Pool Temperatur-e AND

b. Any Table M2ef the, fll*wi* g transient events in progress.

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" EGGS (SI) actuation Month 20XX QC 3-63 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Table M2 Significant Transients

" Turbine Trip

" Reactor Scram

• ECCS Activation

" Recirc. Runback > 25% Reactor Power Change

" Thermal Power oscillations > 10% Reactor Power Change Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product 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 er mereany of I the listed 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). For example, the reactor power level cannot be determined from any analog, computer point, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly 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 action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV-leve!RPV water level and RCS heat removal.

Month 20XX QC 3-64 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY The loss of the ability to determine neo Or mereany of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for ene or-moreany of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted 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 plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RAS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA2 Month 20XX QC 3-65 EP-AA-1006 (Revision XX)

eliinrf r-ifina Annav I=walnn M, r~lakr A ui~rI (~i*iaa Ann~v FvMnn Ma.,,Igimr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSU421 Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 15 minutes-has been exceeded, or will likely be exceeded.

a- An-UNPLANNED event results in the inability to monitor one orFmoRe.ANY Table M1 parameters 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 the fglloWin. Dq...ar.A.t rs, from. Within the Control Room. for 15 m;inuite

• lO II orloge-

2. [BWRpomaranwtc 3. IPWRpa*metcr IiS4 Wi~

4. Reactor Power- 6. Reactor. Pow 57--

8.RVWater- L" &RCS

9. Leve1 10-.RPX' Pressur-e 11. -R-CSFP-r-essur e

12. Prniw 12. In Corv./Come Containment Exit Temp,,rt,-

14. uppessin P15. Levels in at least Leve (site-speetie

17. Steam Cencrator Temper-ataf AffliliniffOf

___ Wate-FNOW Month 20XX QC 3-66 EP-AA-1006 (Revision XX)

ula Exallon Nuclear Eeo Ouad Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically 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 Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one OF moreany of the listed 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). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated 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 it significantly 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 action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine eoe or mereany of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one er mereany of the listed parameters are lost, then the ability to determine 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 the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC MSA42-.

Month 20XX QC 3-67 EP-AA-1006 (Revision XX)

rl"nrl ifine Annav FPylnn umlrlnr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis Reference(s):

1. NEI 99-01 Rev 6, SU2 Month 20XX QC 3-68 EP-AA-1006 (Revision XX)

Text

ATTACHMENT 10 DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FOR QUAD CITIES NUCLEAR POWER STATION EP-AA-1 006 Enclosures 0 Enclosure 10A - EAL Comparison Matrix Document 0 Enclosure 10B - EAL Red-Line Basis Document S Enclosure 10C - EAL Basis Document

NEI 99-01 REVISION 6 DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS ATTACHMENT 10 DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FOR QUAD CITIES NUCLEAR STATION m Exelon Generation.

NEI 99-01 Rev 6 Proposed EAL Justification AG1 RG1 Initiating Condition - GENERAL EMERGENCY Initiating Condition: 1'1 No Change M Difference FIDeviation Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem Release of gaseous radioactivity resulting in offsite dose greater than 1,000 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 5,000 mrem thyroid CDE. mRem TEDE or 5,000 mRem thyroid CDE. classification.

Operating Mode Applicability: All Operating Mode Applicability:

2) Added the following to bullet #3" Classification based on effluent monitor 1,2.3,4,5,D readings assumes that a release path to the environment is established." In order Emergency Action Level (EAL): to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) readings assumes that a release path to the environment is established. If the Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for

• The Emergency Director should declare the General Emergency promptly classification purposes." This allows for more timely classification since all the upon determining that the applicable time has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded.

that the applicable time has been exceeded, or will likely be exceeded. readily available on the 11x17 procedure matrix used by the SM.

• If an ongoing release is detected and the release start time is unknown,

• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes. assume that the release duration has exceeded 15 minutes.

• If the effluent flow past an effluent monitor is known to have stopped due to

• Classification based on effluent monitor readings assumes that a release path actions to isolate the release path, then the effluent monitor reading is no to the environment is established. If the effluent flow past an effluent monitor is longer valid for classification purposes. known to have stopped due to actions to isolate the release path, then the

• The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.

used for emergency classification assessments until the results from a dose

• The pre-calculated effluent monitor values presented in EAL #1 should be assessment using actual meteorology are available.

used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer:

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCilsec for > 15 minutes (as determined by Control Room Panels or PPDS

- Total Noble Gas Release Rate).

(site specific monitor list and threshold values)

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the

2. Dose assessment actual meteorology indicates doses greater than 1000 site boundary of EITHER:

mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose receptor point) a. > 1000 mRem TEDE OR

3. Field survey results indicate EITHER of the following at or beyond (site b. > 5000 mRem CDE Thyroid specific dose receptor point): OR

• Closed window dose rates greater than 1000 mR/hr expected to 3. Field survey results at or beyond the site boundary indicate EITHER:

continue for 60 minutes or longer.

a. Gamma (closed window) dose rates >1000 mR/hr are expected to

• Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.

5000 mrem for one hour of inhalation.

OR

b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

Page I of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification AS1 RS1 Initiating Condition - SITE AREA EMERGENCY Initiating Condition: D No Change I Difference 1-- Deviation Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem Release of gaseous radioactivity resulting in offsite dose greater than 100 mRem 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 500 mrem thyroid CDE. TEDE or 500 mRem thyroid CDE. classification.

Operating Mode Applicability: All Operating Mode Applicability:

2) Added the following to bullet #3" Classification based on effluent monitor 1,2,3,4,5,D readings assumes that a release path to the environment is established." In order Emergency Action Level (EAL): to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) readings assumes that a release path to the environment is established. If the Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for

• The Emergency Director should declare the Site Area Emergency promptly classification purposes." This allows for more timely classification since all the upon determining that the applicable time has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded. that the applicable time has been exceeded, or will likely be exceeded. readily available on the 1 1x17 procedure matrix used by the SM.

• If an ongoing release is detected and the release start time is unknown, I an ongoing release is detected and the release start time is unknown, If assume that the release duration has exceeded 15 minutes.

assume that the release duration has exceeded 15 minutes.

• If the effluent flow past an effluent monitor is known to have stopped due to

• Classification based on effluent monitor readings assumes that a release path actions to isolate the release path, then the effluent monitor reading is no to the environment is established. If the effluent flow past an effluent monitor is longer valid for classification purposes.

known to have stopped due to actions to isolate the release path, then the

• The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.

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

for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer:

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS -

Total Noble Gas Release Rate).

(site specific monitor list and threshold values)

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the

2. Dose assessment actual meteorology indicates doses greater than 1000 site boundary of EITHER:

mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose receptor point) a. > 100 mRem TEDE OR

3. Field survey results indicate EITHER of the following at or beyond (site b. > 500 mRem CDE Thyroid specific dose receptor point):

OR

• Closed window dose rates greater than 100 mR/hr expected to 3. Field survey results at or beyond the site boundary indicate EITHER:

continue for 60 minutes or longer.

a. Gamma (closed window) dose rates >100 mR/hr are expected to

" Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.

500 mrem for one hour of inhalation.

OR

b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.

Page 2 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification AA1 RA1 Initiating Condition - ALERT Initiating Condition:

H No Change Difference L- Deviation Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 1) Listed site-specific monitors and Threshold values to ensure timely classification.

mrem TEDE or 50 mrem thyroid CDE. mrem TEDE or 50 mrem thyroid CDE.

Operating Mode Applicability: All Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1,2,3,4, 5, D to delete the following from the basis "Classification based on effluent monitor Emergency Action Level (EAL): readings assumes that a release path to the environment is established. Ifthe Example Emergency Action Levels: (1 or 2 or 3) effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." This allows for more timely classification since all the basis

• The Emergency Director should declare the Alert promptly upon information pertaining to Note bullet 3 will be contained in the IC and therefor determining that the applicable time has been exceeded, or will likely be

• The Emergency Director should declare the event promptly upon determining readily available on the 11x17 procedure matrix used by the SM.

exceeded. 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,

• If an ongoing release is detected and the release start time is unknown, assume 3) A calculation was performed, in accordance with (IAW) guidance provided in NEI assume that the release duration has exceeded 15 minutes. 99-01 revision 6 EAL AA1, to determine the effluent monitor response for a that the release duration has exceeded 15 minutes.

radioactive liquid release via the normal site release pathway. The liquid release

• If the effluent flow past an effluent monitor is known to have stopped due would contain activity equivalent to provide 10mrem TEDE or 50mrem thyroid CDE Classification based on effluent monitor readings assumes that a release path to actions to isolate the release path, then the effluent monitor reading is to the environment is established. If the effluent flow past an effluent monitor is at the site boundary. The calculation determined the effluent monitor response no longer valid for classification purposes. known to have stopped due to actions to isolate the release path, then the would be >110% of the instrument maximum range and as such, lAW NEI 99-01

• The pre-calculated effluent monitor values presented in EAt #1 should be effluent monitor reading is no longer valid for classification purposes. Rev 6 guidance, was not included in this EAL.

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

1. Reading on any of the following radiation monitors greater than the

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 reading shown for 15 minutes or longer: uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS

- Total Noble Gas Release Rate).

OR (site-specific monitor list and threshold values)

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

2. Dose assessment actual meteorology indicates doses greater than 10

a. > 10 mRem mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose TEDE receptor point)

OR

3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem b. > 50 mRem CDE Thyroid thyroid CDE at or beyond (site-specific dose receptor point) for one hour of OR exposure.

3. Analysis of a liquid effluent sample indicates a concentration or release rate

4. Field survey results indicate EITHER of the following at or beyond (site that would result in doses greater than EITHER of the following at or beyond specific dose receptor point): the site boundary

• Closed window dose rates greater than 10 mR/hr expected to a. 10 mRem TEDE for 60 minutes of exposure continue for 60 minutes or longer.

OR

• Analysis of field survey samples indicate thyroid CDE greater than b. 50 mRem CDE Thyroid for 60 minutes of exposure 50 mrem for one hour of inhalation.

OR

4. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates > 10 mRihr are expected to continue fork 60 minutes.

OR

b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.

Page 3 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification AU1 RUI Initiating Condition - UNUSUAL EVENT Initiating Condition: 1 No Change E Differencee Deviation Release of gaseous or liquid radioactivity greater than 2 times the (site-specific Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 1) Listed site-specific monitors and Threshold values to ensure timely classification.

effluent release controlling document) limits for 60 minutes or longer 60 minutes or longer.

Operating Mode Applicability: Alt Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1,2,3,4,5, D to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for classification purposes." This allows for more timely classification since all the basis

• The Emergency Director should declare the Unusual Event promptly upon

• The Emergency Director should declare the event promptly upon determining information pertaining to Note bullet 3 will be contained in the IC and therefor determining that 60 minutes has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded.

readily available on the 11x17 procedure matrix used by the SM.

exceeded.

• If an ongoing release is detected and the release start time is unknown, assume

• If an ongoing release is detected and the release start time is unknown, that the release duration has exceeded 60 minutes.

assume that the release duration has exceeded 60 minutes.

• Classification based on effluent monitor readings assumes that a release path

• If the effluent flow past an effluent monitor is known to have stopped, to the environment is established. If the effluent flow past an effluent monitor is indicating that the release path is isolated, the effluent monitor reading is known to have stopped due to actions to isolate the release path, then the no longer valid for classification purposes. effluent monitor reading is no longer valid for classification purposes.

1. Reading on ANY effluent radiation monitor greater than 2 times the (site- 1. Reading on ANY of the following effluent monitors > 2 times alarm specific effluent release controlling document) limits for 60 minutes or setpoint established by a current radioactive release discharge permit longer: for a 60 minutes.

(site-specific monitor list and threshold values corresponding to 2 times

• Radwaste Effluent Monitor 1/2-1799-01 the controlling document limits)

OR

• Discharge Permit specified monitor

2. Reading on ANY effluent radiation monitor greater than 2 times the OR alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCilsec for > 60 minutes (as determined by Control Room Panels or PPDS

- Total Noble Gas Release Rate).

3. Sample analysis for a gaseous or liquid release indicates a concentration OR or release rate greater than 2 times (site-specific effluent release controlling document limits) for 60 minutes or longer. 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.

Page 4 of 66

NEI 99-01 Rev 6 Proposed EAL Justification AG2 RG2 No Change Difference Deviation Initiating Condition - GENERAL EMERGENCY Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer. 1) EAL not used in accordance with the discussion in Section 1.4, NRC Order EA-1 2-051, it is recommended that this EAL be implemented when the enhanced Operating Mode Applicability: All spent fuel pool level instrumentation is available for use. The completion of the enhanced SFP level indicators and need for the inclusion of this EAL is being Example Emergency Action Levels: tracked in accordance with Exelon Generation Company, LLC's Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for NOTES: The Emergency Director should declare the General Emergency Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated promptly upon determining that 60 minutes has been exceeded, October 25,2012.

or will likely be exceeded

1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer.

Page 5 of 66

NEI 99-01 Rev 6 Proposed EAL Justification AS2 RS2 Initiating Condition - SITE AREA EMERGENCY AR No Change 1X- Difference []Deviation Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) 1) EAL not used in accordance with the discussion in Section 1.4, NRC Order EA-1 2-051, it is recommended that this EAL be implemented when the enhanced Operating Mode Applicability: All spent fuel pool level instrumentation is available for use. The completion of the enhanced SFP level indicators and need for the inclusion of this EAL is being Example Emergency Action Levels: tracked in accordance with Exelon Generation Company, LLC's Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for

1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated description) October 25,2012.

Page 6 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition - ALERT I Condition: RA2 D No Change E Difference F-- 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: Ail Operating Mode Applicability:

2) EAL #3 not used in accordance with the discussion in Section 1.4, NRC Order 1, 2, 3, 4, 5, D EA-1 2-051, it is recommended that this EAL be implemented when the enhanced Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): spent fuel pool level instrumentation is available for use. The completion of the enhanced SFP level indicators and need for the inclusion of this EAL is being

1. 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 Status OR Report to March 12, 2012 Commission Order Modifying Licenses with Regard for

2. Damage to irradiated fuel resulting in a release of radioactivity from the Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated fuel 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 R1 Radiation Monitor reading >1000 mRem/hr.

(site-specific listing of radiation monitors, and the associated readings, Table R1 setpoints and/or alarms) Fuel Handling Incident Radiation Monitors

3. Lowering of spent fuel pool level to (site-specific Level 2 value).

• 1(2) 1705-16A Fuel Pool Rad Monitor

. 1(2) 1705-16B Fuel Pool Rad Monitor Page 7 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification AU2 RU2 Initiating Condition: UNUSUAL EVENT Initiating Condition: E No Change 1 Difference I: Deviation UNPLANNED loss of water level above irradiated fuel. UNPLANNED loss of water level above irradiated fuel 1) Listed site specific level indication and monitors to ensure timely classification.

Operating Mode Applicability: All Operating Mode Applicability:

1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

1. a. UNPLANNED water level drop in the REFUELING PATHWAY as 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by indicated by ANY of the following: ANY of the following:

" Refueling Cavity water level < 282 in. (Upper Wide range (site-specific level indications).

simulated signal).

OR AND

" Spent Fuel Pool water level < 19 ft. above the fuel (< -4 ft.

indicated level).

OR

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors. " Indication or report of a drop in water level in the REFUELING PATHWAY.

AND (site-specific list of area radiation monitors)

b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R1.

Table RI Fuel Handling Incident Radiation Monitors

• 1(2) 1705-16A Fuel Pool Rad Mon

• 1(2) 1705-16B Fuel Pool Rad Mon Page 8 of 66

Rev 66 55-01 Rev Proposed EAL NEI 99-01 NEI Proposed EAL Justification AA3 RA3 Initiating Condition - ALERT Initiating Condition: No Change 1 Difference II7 Deviation Radiation levels that impede access to equipment necessary for normal plant Radiation levels that impede access to equipment necessary for normal plant operations, 1) Listed site specific plant rooms and areas with identified mode applicability to operations, cooldown or shutdown. cooldown or shutdown. ensure timely classification.

Operating Mode Applicability: All Operating Mode Applicability:

1,2,3,4, 5,D Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

Note: If the equipment in the listed room or area was already inoperable, or Note:

out of service, before the event occurred, then no emergency If the equipment in the room or area listed in Table R3 was already inoperable, or classification is warranted out of service, before the event occurred, then no emergency classification is

1. Dose rate greater than 15 mR/hr in ANY of the following areas: warranted.

• Control Room

• Central Alarm Station 1. Dose rate> 15 mR/hr in ANY of the following Table R2 areas:

• (other site-specific areas/rooms) Table R2 Areas Requiring Continuous Occupancy

2. An UNPLANNED event results in radiation levels that prevent or

• Main Control Room (Unit 1 ARM Station #22) significantly impede access to any of the following plant rooms or

• Central Alarm Station - (by survey) areas:

(site-specific list of plant rooms or areas with entry-related mode OR applicability identified)

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:

Table R3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building* Mode 3 and 4 Turbine Building* Mode 3

• Areas required to establish shutdown cooling Page 9 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU3 Initiating Condition: RU3 [Z No Change Difference Deviation Initiating Condition: UNUSUAL EVENT 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 to Power Operation, Startup, Hot Standby, Hot Shutdown 1.2, 3 maintain consistency with current and previous revisions of Exelon EALs. This will Example Emergency Action Levels: Emergency Action Level (EAL): ensure a timely classification since the threshold values are more aligned with the radiological category vice system category.

1. (Site-specific radiation monitor) reading greater than (site-specific value). 1. Offgas system radiation monitor HI-HI alarm.

OR OR

2. Sample analysis indicates that a reactor coolant activity value is greater than 2. Specific coolant activity > 4.0 uCil/gm Dose equivalent 1-131.

an allowable limit specified in Technical Specifications.

Page 10 of 66

NEI 99-01 rev 6 Fission Product Barrier Matrix FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss { L Potential Loss Loss L Potential Loss 4Loss Potential Loss A. UNPLANNED rapid drop in A. Primary containment pressure

• RCS ActivityI primary containment pressure Primary greater than (site-specific value)

A. (Site specific indications that following primary containment Containment A. Primary containment pressure OR reactor coolant activity is None pressure rise Pressure / greater than (site-specific None B. (site-specific explosive mixture) greater than 300 uCilgm OR Primary value) due to RCS leakage. exists inside primary containment dose equivalent 1-131) B. Primary containment pressure Containment OR Conditions response not consistent with LOCA conditions. C. HCTL exceeded.

cannot be A. RPV water level cannot be RPV water A. restored andlevel maintained above A. Primary containment flooding restored and maintained above

2. RPV Water required. (site-specific RPV water level (site-specific RPV water level None None A. Primary containment flooding Level corresponding to top of active corresponding to the top of required.

fuel) or cannot be determined. active fuel) or cannot be determined.

A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR primary system A. UNISOLABLE break in any of the A. UNISOLABLE fUollowing: ((britepecfic following: site-specificanysthems systems leakage that EITHER of theresults in exceeding folwn:venting B. Intentional per primary EOPs containment 3.RCS Leak with potential for high-energy linefollowing:

Rate/ Primary None None breas) 1. Max Normal Operating OR None Containment OR Temperature C. UNISOLABLE primary system Isolation Failure B. Emergency RPV OR leakage that results in exceeding

2. Max Normal Operating Area EITHER of the following:

Depressurization Radiation Level. 1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

4.Primary A. Primary Containment A. Primary Containment Radiation A. Primary Containment Radiation Containment Radiation Monitor reading None Monitor reading greater than None None Monitor reading greater than (site-Radiation greater than (site-specific (site-specific value). specific value).

value).

5. Emergency A. Any Condition in the opinion A. Any Condition in the opinion of 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 Director of the Emergency Director the Emergency Director that the Emergency Director that E g Directr ta indicatestha inicate Emergency Director that indicates dgment that indicates Loss of the indicates Potential Loss of the indicates Loss of the RCS Potential Loss of the RCS BarrierP Loss of the Containment Barrier.

Judgment Fuel Clad Barrier. Fuel Clad Barrier. Barrier. P Barrier.

Page t1 of 66

Proposed Fission Product Barrier Matrix FGI Loss of any two barriers AND Loss or Potential Loss of third barrier. TED I FSI Loss or Potential Loss of ANY two barriers. IFAI ANY Loss or ANY Potential Loss of either Fuel Clad or RCS fFli]

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity> 300 uCllgm Dose None None

1. RCS Activity Equivalent 1-131. None None None

2. RPV water level cannot be restored and 1. RPV water level cannot be restored and

2. RPV Water 1. Plant conditions indicate Primary maintained 0 -142 inches (TAF) maintained > -142 Inches (TAF) Plant conditions indicate Primary Containment Containment fending is required. OR OR None None flooding is required.

Level 3. RPV water levelcannot be determined. 2. RPV water level cannot be determined.

3. Primary Containment pressure > 56 peig.

OR

1. UNPLANNED rapid drop in Drywell 4. a. Drywellor torus hydrogen concentration

1. Drywellpressure >2.5 peig. pressure followingprimary containment 06%.

3. Primary AND pressure rise. AND Containment None None 2. Drywellpressure rise is due to RCS None OR b. Drywellor tows oxygen concentration Pressure/Conditions leakage 2. Drywellpressure response not consistent a 5%.

withLOCAconditions. OR

5. Heat Capacity Limit(lGA200. Figure M) exceeded.

3. UNISOLABLE primarysystem leakage that

1. UNISOLABLE Main Steam Line(MSL), results in EITHERofthe following:

HPCI,Feedwater, RWCU or RCICline a. Secondary Containment area break. temperature s QGA300 Maximum 4.RCS Leak Rate None None OR Normal None None Emergency RPV Depressurization is

2. required. OR operating levels.

b. Secondary Containment level> OGA area radiation 300 Maximum Normal operating level.

5.Primary Drywellradiation monitor reading Drywellradiation monitor reading Drywellradiation monitor reading Containment > 6.65 E*02R/hr. None I 1O0R/hr (10.00 E+02 R/hr). None None > 1.55 E+03Rbhr Radiation

1. UNISOLABLE direct downstream pathway to the environment exists after pnmary containment isolation signal.

OR

2. Intentional Primary Containment venting/purging per EOPs or SAMGsdue to accident conditions.

6.Primary OR Containment None None None None 3. UNISOLABLE primary system leakage that None Isolation Failure results in EITHERof the following:

a. Secondary Containment area temperature s QGA300, Maximum Safe operating levels.

OR

b. Secondary Containment area radiation level s QGA300, Maximum Safe operating levels

7. Emergency A. AnyCondition in the opinion of the A.AnyCondition in the opinion of the A.ANYCondition in the opinion of the A.Any Condition in the opinion of the A.Any Condition in the opinion of the A.AnyCondition in the opinion of the Emergency Director that indicates Loss Emergency Director that indicates Potential Emergency Director that indicates Loss of Emergency Directorthat indicates Potential Emergency Directorthat indicates Loss ot the Emergency Director that indicates Potential Loss Director Judgment of the Fuel Clad Barrier. Loss ofthe Fuel Clad Barrier the RCS Barrier. Loss of the RCS Barrier. Containment Barrier. of the Containment Barrier.

Page 12 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FCI Category: Fuel Clad Barrier No Change Difference 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. (Site specific indications that reactor coolant activity is greater than 300 uCi/gm Coolant activity > 300 uCilgm Dose Equivalent 1-131.

dose equivalent 1-131)

Page 13 of 66

NEI 99-01 Rev 6 Proposed EAL Justification FC2 Category: Fuel Clad Barrier FC2 Z No Change Differce Deviation Category: Fuel Clad Barrier 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary containment flooding required. 1. Plant conditions indicate Primary Containment flooding is required.

Potential Loss Potential Loss A. RPV water level cannot be restored and maintained above (site-specific RPV 2. RPV water level cannot be restored and maintained > -142 inches (TAF) water level corresponding to top of active fuel) or cannot be determined. OR

3. RPV water level cannot be determined.

Page 14 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC4 Category: Fuel Clad Barrier 'C5 [fl 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 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 6.65 E+02 R/hr.

value).

Page 15 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC6 Category: Fuel Clad BarrierFC No Change Differ Deviation Emergency Director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. 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 of the Fuel Clad Barrier. the Fuel Clad Barrier.

Potential Loss Potential Loss A. 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

NEI 99-01 Rev 6 Proposed EAL Justification Category: Reactor Coolant System Barrier RC1 Category: Reactor Coolant System Barrier RC' E 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary containment pressure greater than (site-specific value) due to RCS 1. Drywell pressure >2.5 psig.

leakage. AND

2. Drywell pressure rise is due to RCS leakage.

Page 17 of 66

NEI 99-01 Rev 6 Proposed EAL Justification RC2 RC2 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier No Change 1 Difference 1 Deviation RPV Water Level RPV Water Level 1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A.RPV water level cannot be restored and maintained above (site-specific RPV 1. RPV water level cannot be restored and maintained > -142 inches (TAF).

level corresponding to the top of active fuel) or cannot be determined. OR

2. RPV water level cannot be determined.

Page 18 of 66

NEI 99-01 Rev 6 Proposed EAL Justification RC3 Category: Reactor Coolant System Barrier RC E No Change Difference Deviation Category: Reactor Coolant System Barrier 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. UNISOLABLE break in any of the following: ( site-specific systems with potential 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater,, RWCU or RCIC for high-energy line breas) line break.

OR OR B. Emergency RPV Depressurization 2. Emergency RPV Depressurization is required.

Potential Loss Potential Loss A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following: 3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature > QGA 300 Maximum Normal

1. Max Normal Operating Temperature. operating levels.

OR OR

2. Max Normal Operating Area Radiation Level. b. operating Containment area radiation level > QGA 300 Maximum Normal Secondarylevels.

Page 19 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Reactor Coolant System Barrier RC4 Category: Reactor Coolant System Barrier 'C5 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 100 R/hr (>1.OOE+02 R/hr).

value).

Page 20 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification RC6 RC7 NOCag [ Dfeee ]Deito Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier No Change Difference Deviation Emergency Director Judgment Emergency director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1, 2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. 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 of the RCS Barrier. the RCS Barrier.

Potential Loss Potential Loss A. 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 of66

NEI 99-01 Rev 6 Proposed EAL Justification CTI Category: Containment Barrier No Change Diff n Deviation Category: Containment Barrier 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 in Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 pressure where pressure is now lowering, the risk of a potential loss of containment is no longer present, this wording is also consistent with present EAL Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: wording.

Loss Loss C. UNPLANNED rapid drop in primary containment pressure following primary 1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.

containment pressure rise OR OR 2. Drywell pressure response not consistent with LOCA conditions.

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

3. Drywall pressure >56 psig and rising.

D. Primary containment pressure greater than (site-specific value) OR OR 4. a. Drywell or torus Hydrogen concentration > 6%.

E. (site-specific explosive mixture) exists inside primary containment AND OR b. Drywell or torus Oxygen concentration > 5%.

OR

3. HCTL exceeded. 5. Heat Capacity Limit (QGA 200, Fig. M) exceeded.

Page 22 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT2 Category: Containment Barrier CT2 No Change Difference Deviation Category: Containment Barrier RPV Water Level RPV Water Level Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss - Potential Loss A. Primary containment flooding required. Plant conditions indicate Primary Containment flooding is required.

Page 23 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT3 CT6 Category: Containment Barrier Category: Containment Barrier E No Change 1 Difference 1 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 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss LoSS A. UNISOLABLE direct downstream pathway to the environment exists after 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal primary containment isolation signal.

OR OR

2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to B. Intentional primary containment venting per EOPs accident conditions.

OR OR

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

C. UNISOLABLE primary system leakage that results in exceeding EITHER of the

a. Secondary Containment area temperature > OGA 300 Maximum Safe following:

operating levels.

1. Max Safe Operating Temperature. OR OR b. Secondary Containment area radiation level > QGA 300 Maximum Safe operating levels.

2. Max Safe Operating Area Radiation Level.

Page 24 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Containment Barrier CT4 Category: Containment BarrierT No Change Differnc 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, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss Potential Loss A. Primary Containment Radiation Monitor reading greater than (site-specific value). Dryweil radiation monitor reading > 1.55 E+03 R/h.

Page 25 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT6 Category: Containment BarrierCTI No Change Diffrenc. Deviation Category: Containment Barrier Emergency director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. 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 of the Containment Barrier. the Containment Barrier.

Potential Loss Potential Loss A. 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

NEI 99-01 Rev 6 j Proposed EAL Justification SG1 MG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition: F No Change FIDifference FIDeviation Prolonged loss of all offsite and all onsite AC power to emergency buses. Prolonged loss of all offsite and all onsite AC power to emergency buses.

1) Listed site specific equipment, site specific time based on station blackout Operating Mode Applicability: Operating Mode Applicability: coping analysis, and site specific indication to ensure timely classification.

1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emergency promptly upon determining that (site-specific hours) has been exceeded, or will Note:

likely be exceeded.

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

1. a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).

1 Loss of ALL offsite AC power to unit ECCS busses.

AND AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel

b. EITHER of the following: generators to supply power to unit ECCS busses.

AND

" Restoration of at least one emergency bus in less than (site-specific hours) is not likely. 3. EITHER of the following:

a. Restoration of at least one unit ECCS bus in <1 hour is not likely.

" (Site-specific indication of an inability to adequately OR remove heat from the core)

b. RPV water level cannot be restored and maintained > -166 inches.

Page 27 of 66

0 NEI 99-01 Rev 6 Proposed EAL [ Justification SS1 MS1 Initiating Condition: SITE AREA EMERGENCY Initiating Condition:

No Change FIDifference F Deviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer.

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

Operating Mode Applicability: Operating Mode Applicability:

1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Emergency Action Level (EAL):

Note: The Emergency Director should declare the Site Area Emergency Note:

promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) exceeded.

for 15 minutes or longer.

1. Loss of ALL offsite AC Power to unit ECCS buses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Page 28 of 66

NEI 99-01 Rev 6 Proposed EAL I Justification SA1 MAI Initiating Condition: ALERT Initiating Condition:

M No Change -- Difference D 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 longer.

1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

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

The Emergency Director should declare the event promptly upon

1. a. AC power capability to (site-specific emergency buses) is reduced to a determining that the applicable time has been exceeded, or will likely be single power source for 15 minutes or longer. exceeded.

AND

1. AC power capability to unit ECCS buses reduced to only one of the following

b. Any additional single power source failure will result in loss of all AC power sources for> 15 minutes.

power to SAFETY SYSTEMS.

• Reserve auxiliary Transformer TR-12 (TR-22)

• Unit Auxiliary Transformer TR-1 1 (TR-21)

• Unit Emergency Diesel Generator

• Shared Emergency Diesel Generator

• Station Blackout Diesel Generator

• Unit crosstie breakers AND

2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Page 29 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification Sul MU1 Initiating Condition: UNUSUAL EVENT Loss of all offsite AC power capability to emergency buses for 15 minutes or Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or M No Change FIDifference FIDeviation longer. longer. 1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

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

exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 exceeded.

minutes or longer Loss of ALL offsite AC power capability to unit ECCS buses for > 15 minutes.

Page 30 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SG8 Initiating Condition: GENERAL EMERGENCY Loss of all AC and Vital DC power sources for 15 minutes or longer.

Initiating Condition:

MG2 D No Change M Difference FIDeviation Loss of all AC and Vital DC power sources for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability: 2) in EAL #3 removed the word "indicated" this will allow for an indication problem Operating Mode Applicability: to not cause confusion on the need to declare.

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be Note:

exceeded.

1. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be buses) for 15 minutes or longer. exceeded.

AND

1. Loss of ALL offsite AC power to unit ECCS busses.

Indicated voltage is less than (site-specific bus voltage value) on ALL AND (site-specific vital DC buses) for 15 minutes or longer.

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.

AND

3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.

AND

4. ALL AC and Vital DC power sources have been lost for >15 minutes.

Page 31 of 66

NEt 99-01 Rev 6 Proposed EAL I Justification S33 MS2 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: D No Change M Difference I Deviation Loss of all Vital DC power for 15 minutes or longer.

Loss of all Vital DC power for 15 minutes or longer.

1) Listed site specific equipment and value to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability: 2) Removed the word "indicated" this will allow for an indication problem to not Power Operation, Startup, Hot Standby, Hot Shutdown cause confusion on the need to declare.

1,2,3 Example Emergency Action Levels:

Emergency Action Level (EAL):

Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:

Indicated voltage is less than (site-specific bus voltage value) on ALL Vital DC The Emergency Director should declare the event promptly upon buses for 15 minutes or longer. determining that the applicable time has been exceeded, or will likely be exceeded.

Voltage is < 105 VDC on 125 VDC battery buses #1 and #2 for >15 minutes..

Page 32 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification SS5 Initiating Condition: SITE AREA EMERGENCY Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPV Initiating Condition:

MS3 D1 No Change M Difference

• Deviation water level [BWR]) or RCS heat removal. Inability to shutdown the reactor causing a challenge to RPV water level or RCS.

heat removal. 1) Listed site specific indications to ensure timely classification.

Operating Mode Applicability: 2) Mode 2 included in operating mode applicability as per developer notes.

Operating Mode Applicability:

Power Operation 1,2 Example Emergency Action Levels:

Emergency Action Level (EAL):

1 a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the reactor.

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power

> 5%.

AND AND

b. All manual actions to shutdown the reactor have been unsuccessful. 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%'

AND AND

c. EITHER of the following conditions exist: 3. EITHER of the following conditions exist:

1. (Site-specific indication of an inability to adequately " RPV water level cannot be restored and maintained > -166 inches.

remove heat from the core) OR

" Heat Capacity Limit (QGA 200, Figure M) exceeded.

OR

2. (Site-specific indication of an inability to adequately remove heat from the RCS)

Page 33 of 66

0 NEI 99-01 Rev 6 1 Proposed EAL Justification SAO MA3 Initiating Condition: ALERT Initiating Condition: FH1 No Change M Difference FIDeviation Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and Automatic or manual scram fails to shutdown the reactor, and subsequent manual 1) Listed site specific indications to ensure timely classification.

subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. actions taken at the reactor control consoles are not successful in shutting down the reactor. 2) Mode 2 included in operating mode applicability as per developer notes.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation 1,2 Example Emergency Action Levels:

Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually Note:

driving in control rods or implementation of boron injection strategies.

A manual action is any operator action, or set of actions, which causes the

1. a. An automatic (trip [PWR] / scram [BWRJ) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor. manually driving in control rods or implementation of boron injection strategies.

AND

1. Automatic or manual scram did not shutdown the reactor as indicated by

b. Manual action taken at the reactor control consoles are not Reactor Power > 5%.

successful in shutting down the reactor. AND

2. Manual I ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.

Page 34 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU5 MU3 Initiating Condition: UNUSUAL EVENT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor.

Initiating Condition: FH No Change J--M Difference '- Deviation Automatic or manual scram fails to shutdown the reactor.

1) Listed site specific indications to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

2) Mode 2 included in operating mode applicability as per developer notes.

Power Operation 1,2 Example Emergency Action Levels: (1 or 2)

Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note:

A manual action is any operator action, or set of actions, which causes the

1. a. An automatic (trip [PWR] I scram [BWRJ) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor. manually driving in control rods or implementation of boron injection strategies.

AND

1. a. Automatic scram did not shutdown the reactor as indicated by Reactor

b. A subsequent manual action taken at the reactor control consoles is Power > 5%

successful in shutting down the reactor.

AND

b. Subsequent manual / ARI action taken at the Reactor Console is

2. a. A manual scram ([PWR] / scram [BWR]) did not shutdown the successful in shutting down the reactor.

reactor.

OR AND 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power > 5%.

b. EITHER of the following: AND

b. EITHER of the following:

1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. 1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.

OR OR

2. Subsequent automatic scram / ARI is successful in shutting down

2. A subsequent automatic (trip [PWR] / scram [BWR]) is successful the reactor.

in shutting down the reactor.

Page 35 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SA2 MA4 Initiating Condition: ALERT Initiating Condition: F No Change IDifference FI Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Note: The Emergency Director should declare the Alert promptly upon Emergency Action Level (EAL):

determining that 15 minutes has been exceeded, or will likely be exceeded.

Note:

The Emergency Director should declare the event promptly upon

1. a. An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.

[see table below]

1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter

[ BVVR parameter list] [PWR parameter list] from within the Control Room for >15 minutes.

Reactor Power Reactor Power Table M1 Control Room Parameters RPV Level RCS Level

" Reactor Power

• RPV Water Level RPV Pressure RCS Pressure

• RPV Pressure

" Drywell Pressure Primary Containment Pressure In Core/Core Exit Temperature " Torus Level

" Torus Temperature Levels in at least (site specific Suppression Pool Level number) steam generators AND 1 nPool Temperature Steam Generator Auxiliary or

2. ANY Table M2 transient in progress.

Suppression PEmergency Feed Water Flow AND Table M2 Significant Transients

b. Any of the following transient events in progress. " Turbine Trip

• Automatic or Manual runback greater than 25% thermal reactor power " Reactor Scram

• Electrical load rejection greater than 25% full electrical load " ECCS Activation

• Reactor Scram [BWR] / trip [PWR]

• ECCS (SI) actuation " Recirc. Runback > 25% Reactor Power Change

• Thermal power oscillations greater than 10% [BWR]

• Thermal Power oscillations > 10% Reactor Power Change Page 36 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU2 MU4 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change FIDifference FIDeviation UNPLANNED loss of Control Room indications for 15 minutes or longer.

UNPLANNED loss of Control Room indications for 15 minutes or longer.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:

Note: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):

determining that 15 minutes has been exceeded, or will likely be Note:

exceeded.

The Emergency Director should declare the event promptly upon An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.

[see table below] UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for > 15 minutes.

[ BWVRparameter list] [PWR parameter list]

Table M1 Control Room Parameters Reactor Power Reactor Power

• Reactor Power RPV Level RCS Level

• RPV Water Level

• RPV Pressure

• Drywell Pressure RPV Pressure RCS Pressure

• Torus Level

• Torus Temperature Primary Containment Pressure In Core/Core Exit Temperature Levels in at least (site specific Suppression Pool Level number) steam generators Steam Generator Auxiliary or Suppression Pool Temperature Emergency Feed Water Flow Page 37 of 66

NEI 99-01 Rev 6 j Proposed EAL Justification SA9 MA5 Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM needed for the current operating Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM required for the current operating FH No Change FI Difference H Deviation

1) No additional site specific hazard noted mode. mode.

2) Changed the word "needed" to "required" in the IC and "required by Technical Operating Mode Applicability: Operating Mode Applicability: Specification" in the EAL to be consistent with terminology used by operators and minimize confusion.

1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Level (EAL):

Example Emergency Action Levels:

1. The occurrence of ANY of the following hazardous events:

1. a. The occurrence of ANY of the following hazardous events: " Seismic event (earthquake)

• Internal or external flooding event

• Seismic event (earthquake)

• Internal or external flooding event

• High winds or tornado strike

• High winds or tornado strike

• FIRE

• FIRE " EXPLOSION

• EXPLOSION

" Other events with similar hazard characteristics as determined

• (site-specific hazards) by the Shift Manager

• Other events with similar hazard characteristics as determined by the Shift Manager AND

2. EITHER of the following:

AND a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM

b. EITHER of the following: required by Technical Specifications for the current operating mode.

1. Event damage has caused indications of degraded performance OR in at least one train of a SAFETY SYSTEM needed for the current operating mode. b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical OR Specifications for the current operating mode.

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

Page 38 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU4 MU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change M Difference FIDeviation RCS leakage for 15 minutes or longer.

RCS leakage for 15 minutes or longer. 1) Listed site specific values to ensure timely classification.

Operating Mode Applicability: 2) Changed wording from containment to Drywell for clarity to better define the Operating Mode Applicability: primary containment structure.

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 3) In EAL #1 and 2 added "into the Drywell" to differentiate between EAL #1/2 and

1. 3. Without this wording would have been in EAL #1 or #2 concurrent with #3. With Example Emergency Action Levels: (1 or 2 or 3) the added wording each EAL can be called separately.

Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unusual Event promptly upon Note:

determining that 15 minutes has been exceeded, or will likely be exceeded.

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

1. RCS unidentified or pressure boundary leakage greater than (site-specific exceeded.

value) for 15 minutes or longer.

1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for

2. RCS identified leakage greater than (site-specific value) for 15 minutes or > 15 minutes.

longer OR

2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.

3. Leakage from the RCS to a location outside containment greater than 25 OR gpm for 15 minutes or longer

3. Leakage from the RCS to a location outside the Drywell >25 gpm for

> 15 minutes.

Page 39 of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification sub MU7 Initiating Condition: UNUSUAL EVENT Loss of all onsite or offsite communications capabilities Initiating Condition: M No Change FIDifference FIDeviation Loss of all onsite or offsite communication capabilities. 1) Listed site specific communication methods to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels: (1 or 2 or 3)

Emergency Action Level (EAL):

1. Loss of ALL of the following onsite communication methods:

(site-specific list of communications method 1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine operations.

2. Loss of ALL of the following ORO communications s) methods: OR

2. Loss of ALL Table M3 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.

OR

3. Loss of ALL of the following NRC communications methods:

3. Loss of ALL Table M3 NRC communication capability affecting the ability to perform NRC notifications.

(site-specific list of communications methods)

Table M3 Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X Sound Powered Phones X All telephone Lines (Commercial and X X X microwave)

ENS X X HPN X X Satellite Phones X X Page 40 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA2 CAI Initiating Condition: ALERT Initiating Condition: E No Change 1 1 Difference [: Deviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.

longer.

Operating Mode Applicability:

Operating Mode Applicability:

4,5, D Cold Shutdown, Refueling, Defueled Emergency Action Level (EAL):

Example Emergency Action Levels:

Note:

Note: The Emergency Director should declare the Alert promptly upon The Emergency Director should declare the event promptly upon determining determining that 15 minutes time has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded.

exceeded.

Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 minutes or longer. 1. Loss of ALL offsite AC power to unit ECCS buses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Page 41 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: C No Change Difference Deviation Initiating Condition: UNUSUAL EVENT 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.

longer.

Operating Mode Applicability:

Cold Shutdown, Refueling, Defueled Operating Mode Applicability:

4, 5, D Example Emergency Action Levels:

Note: 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 single power source for 15 minutes or longer, determining that the applicable time has been exceeded, or will likely be exceeded.

AND

b. Any additional single power source failure will result in loss of all AC 1. AC power capability to unit ECCS buses reduced to only one of the power to SAFETY SYSTEMS. following power sources for > 15 minutes.

• Reserve auxiliary Transformer TR-12 (TR-22)

• Unit auxiliary transformer TR-1 1 (TR-21)

• Unit Emergency Diesel Generator

• Shared Emergency Diesel Generator

• Station Blackout Diesel Generator

• Unit crosstie breakers AND

2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Page 42 of 66

Justification Proposed EAL Rev 6 99-01 Rev NEI 99-01 6 Proposed EAL Justification Initiating Condition - ALERT Hazardous event affecting SAFETY SYSTEM needed for the current operating CA6 Initiating Condition:

Hazardous event affecting SAFETY SYSTEM required for the current operating CA2 H No Change 1 -lDifference Deviation

1) No additional site specific hazards noted mode. mode.

Operating Mode Applicability: Operating Mode Applicability:

2) Changed the word "needed" to "required" in the IC and "required by Technical Specification" in the EAL to be consistent with terminology used by Cold Shutdown, Refueling 4,5 operators and minimize confusion.

Example Emergency Action Levels: Emergency Action Level (EAL):

1. a. The occurrence of ANY of the following hazardous events: 1. The occurrence of ANY of the following hazardous events:

• Seismic event (earthquake)

• Seismic event (earthquake)

• Internal or external flooding event

• Internal or external flooding event

• High winds or tornado strike

• High winds or tornado strike

• FIRE

• FIRE

" EXPLOSION

• EXPLOSION

" (site-specific hazards)

• Other events with similar hazard characteristics as determined by the Shift Manager

" Other events with similar hazard characteristics as determined by the Shift Manager AND AND 2. EITHER of the following:

b. EITHER of the following: a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM

1. Event damage has caused indications of degraded required by Technical Specifications for the current performance in at least one train of a SAFETY SYSTEM operating mode.

needed for the current operating mode.

OR OR

b. The event has caused VISIBLE DAMAGE to a SAFETY

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current SYSTEM component or structure required by technical operating mode. Specifications for the current operating mode.

Page 43 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CU4 CU3 D N hne D Dfeec - eito Initiating 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 not Cold 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 exceeded.

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Indicated voltage is less than (site-specific bus voltage value) on required Vital DC exceeded.

buses for 15 minutes or longer.

Voltage is < 105 VDC on required 125 VDC battery buses #1 and #2 for> 15 minutes.

Page 44 of 66

NEI 99-01 Rev 6 ] Proposed EAL Justification CU5 CU4 Initiating Condition: UNUSUAL EVENT Initiating Condition: 1 -l No Change [: Difference -- Deviation Loss of all onsite or offsite communications capabilities Loss of all onsite or offsite communication capabilities. 1) Listed site specific communications methods to ensure timely classification Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown, Refueling, Defuled 4,5, D Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL):

1. Loss of ALL of the following onsite communication methods: I. Loss of ALL Table C1 Onsite communications capability affecting the ability to perform routine operations.

(site-specific list of communications method OR

2. Loss of ALL of the following ORO communications s) methods:

2. Loss of ALL Table C1 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.

3. Loss of ALL of the following NRC communications methods:

OR (site-specific list of communications methods)

3. Loss of ALL Table C1 NRC communication capability affecting the ability to perform NRC notifications.

Table Cl Communications Capab lity System Onsite Offsite NRC Plant Radio X Plant Page X Sound Powered Phones x All telephone Lines (Commercial and X X X microwave)

ENS x x HPN x x Satellite Phones X X Page 45 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA3 CAS Initiating Condition: ALERT Initiating Condition: M No Change E: Difference 1 Deviation Inability to maintain the plant in cold shutdown. Inability to maintain plant in cold shutdown. 1) Listed site specific Technical Specification cold shutdown temperature limit and site-specific pressure reading to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown, Refueling 4,5 Example 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 exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be

1. UNPLANNED increase in RCS temperature to greater than (site-specific exceeded.

Technical Specification cold shutdown temperature limit) for greater than the duration specified in the following table.

2. UNPLANNED RCS pressure increase greater than (site-specific pressure 1. UNPLANNED rise in RCS temperature > 212?F due to loss of decay reading). (This EAL does not apply during water-solid plant conditions. heat removal for > Table C2 duration.

[PWRJ)

Table C2 RCS Heat-up Duration Thresholds Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration RCS Status Containment Closure Heat-up Duration Status Status Intact Not Applicable 60 minutes*

Intact (but not RCS Reduced Not Applicable 60 minutes* Established 20 minutes*

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

reduced inventory Not Established 0 minutes

[PWR]) Not Established 0 minutes Ifan RCS heat removal system is in operation within this time

• If an RCS heat removal system is in operation within this time frame and frame and RCS temperature is being reduced, then EAL #1 is not RCS temperature is being reduced, the EAL is not applicable. asolicable.

OR

2. UNPLANNED RPV pressure rise >10 psig as a result of temperature rise due to loss of decay heat removal.

Page 46 of 66

0 NEI 99-01 Rev 6 Proposed EAL Justification CU3 CUS E Ne Initiating Condition: UNUSUAL EVENT Initiating Condition: Li No Change Difference Deviation UNPLANNED increase in RCS temperature. UNPLANNED rise in RCS temperature. 1) Listed site specific Technical Specification cold shutdown temperature limit Operating 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 likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be

1. 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 vesselIRCS [PAR] or RPV

[BWR8 ) level indication for 15 minutes or longer,. UNPLANNED rise in RCS temperature> 212F due to loss of decay heat removal.

OR

2. Loss of the following for >15 minutes.

• ALL RCS temperature indications AND

• ALL RPV water level indications Page 47 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: GENERAL EMERGENCY CGI Initiating Condition:

CG6

- No Change Wl Difference -- Deviation Loss of (reactor vessel/RCS (PWRJ or RPV [BWR]) inventory affecting fuel clad Loss of reactor vessel / RCS inventory affecting fuel clad integrity with integrity with containment challenged. containment challenged. 1) Listed site specific levels, radiation monitors, and sumps and tanks to ensure Operating Mode Applicability: timely classification.

Operating Mode Applicability:

Cold Shutdown, Refueling 4,5 2) Listed Explosive mixture in the Containment Challenge Table to ensure timely classification.

Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emergency Note: 3) Worded "cannot be monitored" as unknown to ensure clarity for instances promptly upon determining that 30 minutes time has been exceeded, or when the indicator is working but is over/under ranged. This is also in keeping The Emergency Director should declare the event promptly upon will likely be exceeded. with current EAL wording.

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

1. a. (Reactor vessel/RCS (PWR] or RPV [BWR]) vessel level less than (site- exceeded.

specific level) for 30 minutes or longer.

1. a. RPV water level < -142 inches (TAF) for > 30 minutes.

AND AND

b. ANY indication from the Containment Challenge Table b. Any Containment Challenge Indication (Table C4)

2. a.. (Reactor vessel/RCS EPWR) or RPV [SWR]) vessel level cannot be OR monitored for 30 minutes or longer. 2. a. RPV water level unknown for > 30 minutes.

AND AND

b. Core uncovery is indicated by ANY of the following: b. Core uncovery is indicated by ANY of the following:

• Table C3 indications of a sufficient magnitude to indicate core

• (Site-specific radiation monitor) reading greater than (site-specific value) uncovery.

• Erratic source range monitor indication [PWVR] OR

• UNPLANNED increase in (site-specific sump and/or tank levels) of

• Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.

sufficient magnitude to indicate core uncovery AND

• (Other site-specific indications) c. ANY Containment Challenge Indication (Table C4)

AND Table C3 Indications of RCS Leakage

c. ANY indication from the Containment Challenge Table).

• UNPLANNED floor or equipment sump level rise*

Table: Containment Challenge Table

• UNPLANNED Torus level rise*

• CONTAINMENT CLOSURE not established*

• (Explosive mixture) exists inside containment

• UNPLANNED vessel make up rate rise

• UNPLANNED increase in containment pressure

• Observation of leakage or inventory loss

• Secondary containment radiation monitor reading above (site-specific value) [BWR] 'Rise in level is attributed to a loss of RPV inventory.

• if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required. Table C4 Containment Challenge Indications

• Primary Containment Hydrogen Concentration _>6% and Oxygen

> 5%

• UNPLANNED rise in containment pressure

• CONTAINMENT CLOSURE not established' ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.

'if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.

Page 48 of 66

Justification Proposed EAL NEI 99-01 NEI Rev 66 99-01 Rev Proposed EAL Justification i

CS1 CS6 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: [-1 No Change W Difference E ] Deviation Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting core Loss of reactor vessel / RCS inventory affecting core decay heat removal 1) Listed site specific values for level, radiation monitors, and sumps and tanks decay heat removal capability. capabilities. to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability: 2) EAL #1 Quad Cities Station does not have a low, low, low RPV water level Cold Shutdown, Refueling 4,5 value stated within their Technical Specifications, all ESF functions come from the low, low value of -55.2inches, NEI 99-01 rev 6 developer notes for CS1 Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): request the developer to use the low, low, low value for declaration of a SAE Note: The Emergency Director should declare the Site Area Emergency Note: when Containment Closure is not established, as such an RPV water level -65 promptly upon determining that 30 minutes time has been exceeded, or inches was selected as the threshold value since itis substantially above TAF The Emergency Director should declare the event promptly upon RPV water level of -142 inches and substantially below the low, low value of will likely be exceeded.

determining that the applicable time has been exceeded, or will likely be -55.2inches clearly indicating a worsening of the condition and the need to

1. a. CONTAINMENT CLOSURE not established. exceeded. escalate the emergency from an Alert condition.

AND

. With CONTAINMENT CLOSURE not established, RPV water level 3) Worded "cannot be monitored" as unknown to ensure clarity for instances

b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site- when the indicator is working but is over/under ranged. This is also in keeping specific level). <-65 inches with current EAL wording.

2. a. CONTAINMENT CLOSURE established. OR AND 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 inches (TAF)

b. (Reactor vessel/RCS [PIVR] or RPV [BWVRJ)level less than (site-specific level). OR

3. a. (Reactor vessel/RCS [PWI/R] or RPV [BVVR]) level cannot be 3. a. RPV water level unknown for > 30 minutes monitored for 30 minutes or longer.

AND AND

b. Core uncovery is indicated by ANY of the following:

b. Core uncovery is indicated by ANY of the following:

" Table C3 indications of a sufficient magnitude to indicate core

• (Site-specific radiation monitor) reading greater than (site- uncovery.

specific value)

OR

• Erratic source range monitor indication [PWR]

" Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.

• UNPLANNED increase in (site-specific sump and/or tank levels) of sufficient magnitude to indicate core uncovery

• (Other site-specific indications) 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 49 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA1 CA6 Initiating Condition: ALERT Initiating Condition: D No Change W Difference Deviation Loss of (reactor vessel/RCS [PWRI or RPV [BWVRI)inventory Loss of RPV inventory 1) Listed site specific levels, and sumps and tanks to ensure timely Operating Mode Applicability: Operating Mode Applicability: classification.

Cold Shutdown, Refueling 4,5 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 keeping Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

with current EAL wording.

Note: The Emergency Director should declare the Alert promptly upon Note:

determining that 15 minutes time has been exceeded, or will likely be exceeded.

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

1. Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as exceeded.

indicated by level less than (site-specific level).

2. a. (Reactor vesselIRCS [PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer

1. Loss of RPV inventory as indicated by level < - 59inches.

AND OR

b. UNPLANNED increase in (site-specific sump and/or tank) levels due 2. a. RPV water level unknown for> 15 minutes.

to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. AND

b. Loss of RPV inventory per Table C3 indications.

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

NEI 99-01 Rev 6 Proposed EAL Justification Cu1 CU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change E Difference 1: Deviation UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory for UNPLANNED loss of reactor vessel / RCS inventory for 15 minutes or longer. 1) Described 'a required lower limit" as a procedurally established lower limit, 15 minutes or longer.

Operating Mode Applicability:

and listed site specific sumps and tanks to ensure timely classification.

Operating Mode Applicability:

4,5 Cold Shutdown, Refueling 2) Worded "cannot be monitored" as unknown to ensure clarity for instances Emergency Action Level (EAL):

Example Emergency Action Levels: (1 or 2) when the indicator is working but is over/under ranged. This is also in keeping Note: with current EAL wording.

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

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

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

[PWR] or RPV [BWRI) level less than a required lower limit for 15 minutes or longer. 1. UNPLANNED loss of reactor coolant results in the inability to restore and

2. a. (Reactor vessel/RCS [PAR] or RPV [BWVR])level cannot be maintain RPV water level above the procedurally established lower limit for> 16 minutes.

monitored.

AND OR

b. UNPLANNED increase in (site-specific sump and/or tank) levels. 2. a. RPV water level unknown AND

b. Loss of RPV inventory per Table C3 indications.

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

NEI 99-01 Rev 6 Proposed EAL Justification HGl HGI flf f- ifeene {* Dvito Initiating Condition: GENERAL EMERGENCY Initiating Condition: iG No Change J Difference LJDeviation 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 Operating Mode Applicability: Operating Mode Applicability: timely classification.

All 1,2, 3, 4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift 1. A notification from the Security Force that a HOSTILE ACTION is occurring or has supervision). occurred within the PROTECTED AREA.

AND AND 2. a. ANY Table H1 safety function cannot be controlled or maintained.

b. EITHER of the following: OR

b. Damage to spent fuel has occurred or is IMMINENT

1. ANY of the following safety functions cannot be controlled or maintained.

Table H1 Safety Functions

• Reactivity control . Reactivity Control

• Core cooling [PWR] / RPV water level [BW4R] (ability to shut down the reactor and keep it shutdown)

RCS heat removal . RPV Water Level (ability to cool the core)

• RCS Heat Removal (ability to maintain heat sink)

OR

2. Damage to spent fuel has occurred or is IMMINENT Page 52 of 66

NEI 99-01 Rev 6 Proposed EAL I Justification HSI Initiating Condition: SITE AREA EMERGENCY Initiating Condition:

HSI M No Change 1111] Difference [-] Deviation HOSTILE ACTION within the Protected Area. HOSTILE ACTION within the Protected Area. 1) List site security shift supervision as Security Force.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

A HOSITLE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-security shift supervision). A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

Page 53 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: ALERT Initiating Condition: H No Change I Difference FIDeviation HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack 1) List site security shift supervision as Security Force.

threat within 30 minutes. threat within 30 minutes.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):

1. A HOSTILE ACTION is occurring or has occurred within the OWNER 1 A validated notification from NRC of an aircraft attack threat < 30 minutes CONTROLLED AREA as reported by the (site-specific security shift from the site.

supervision).

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

2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.

Page 54 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HUl HUI Initiating Condition: UNUSUAL EVENT Initiating Condition: [F No Change F-- Difference L Deviation Confirmed SECURITY CONDITION or threat.

Confirmed SECURITY CONDITION or threat. 1) List site security shift supervision as Security Force.

Operating Mode Applicabllity: 2) Further described credible security threat through listing a site specific Operating Mode Applicability: procedure.

All 1, 2, 3,4,5, D Example Emergency Action Levels: (I or 2 or 3)

Emergency Action Level (EAL):

1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision). 1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.

2. Notification of a credible security threat directed at the site. OR

2. A validated notification from the NRC providing information of an aircraft

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

threat.

OR

3. Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.

Page 55 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HS6 HS2mmm Initiating Condition: SITE AREA EMERGENCY Initiating Condition: iS2 No Change I[ ] Difference [ Deviation Inability to control a key safety function from outside the Control Room. Inability to control a key safety function from outside the Control Room. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would Operating Mode Applicability: Operating Mode Applicability: enter should such an event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate All 1,2, 3, 4, 5, D shutdown were to be listed,

2) Added descriptors to better explain each safety function and allow for a Example Emergency Action Levels: (1 and 2) Emergency Action Level (EAL): timely classification.

Note: The Emergency Director should declare the Site Area Emergency Note:

promptly upon determining that (site-specific number of minutes) has 3) Changed "An event" to" A Control Room evacuation" to remove confusion been exceeded, or will likely be exceeded.

• The Emergency Director should declare the event promptly upon determining that if partial plant control was transferred to outside the control room with the the applicable time has been exceeded, or will likely be exceeded. control room still manned, due to testing or equipment failure.

1. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). 1 A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per

2. Control of ANY of the following key safety functions is not reestablished

• QCARP 0050-01, SB-i-1 Injection with SSMP and Bringing the Unit to within (site-specific number of minutes). Cold Shutdown OR Reactivity control . OCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to

• Core cooling [PWR] /RPV water level [BWR] Cold Shutdown

• RCS heat removal OR

. QOA 0010-05, Plant Operation with the Control Room Inaccessible AND

2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.

Table H1 Safety Functions

• Reactivity Control (ability to shut down the reactor and keep it shutdown)

• RPV Water Level (ability to cool the core)

• RCS Heat Removal (ability to maintain heat sink)

Page 56 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HAS H'A2 mmm Initiating Condition: ALERT Initiating Condition: [J No Change [j- Difference FIDeviation Control Room evacuation resulting in transfer of plant control to altemate locations. Control Room evacuation resulting in transfer of plant control to alternate locations. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would enter should such an Operating Mode Applicability: Operating Mode Applicability: event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate shutdown were to be listed, All 1, 2, 3,4,5, 0

2) Changed "An event" to" A Control Room evacuation" to remove confusion if partial plant control was transferred to outside the control room with the control Example 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 has resulted in plant control being transferred from the (site-specific remote shutdown panels and local control stations). A Control Control Room Room to evacuation alternate locations per:

• QCARP 0050-01, SB-i-i Injection with SSMP and Bringing the Unit to Cold Shutdown OR

• QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR

• QOA 0010-05, Plant Operation with the Control Room Inaccessible Page 57 of 66

NEI 99-Cl Rev 6 1 Proposed EAL Justification HU4 HU3 Initiating Condition: UNUSUAL EVENT FIRE potentially degrading the level of safety of the plant.

Initiating Condition: M No Change I Difference FIDeviation FIRE potentially degrading the level of safety of the plant. 1) Listed site specific list of plant rooms or areas that contain SAFETY SYSTEM equipment to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability:

All 1,2,3,4,5, D Example Emergency Action Levels: (1 or 2 or 3 or 4)

Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been Note:

exceeded, or will likely be exceeded.

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

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

• Report from the field (i.e., visual observation) A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of the

• Receipt of multiple (more than 1) fire alarms or indications following FIRE detection indications:

• Field verification of a single fire alarm

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

AND

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

b. The FIRE is located within ANY of the following plant rooms or

• Field verification of a single fire alarm areas: Table H2 Vital Areas (site-specific list of plant rooms or areas)

• Reactor Building (when inerted the Drywell is exempt)

• Main Control Room Envelope

2. a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).

• Unit and Shared Emergency Diesel Generator Rooms AND

• 4KV Switchgear Area

b. The FIRE is located within ANY of the following plant rooms or

• Battery Rooms areas:

• RHR Service Water Vaults (site-specific list of plant rooms or areas)

• Turbine Building Cable Tunnel AND

• Cribhouse

c. The existence of a FIRE is not verified within 30-minutes of alarm OR receipt.

2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of a FIRE).

3. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the AND plant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. b. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.

OR

3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-

4. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the minutes of the initial report, alarm or indication.

plant Protected Area] PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish. OR

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

Page 58 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT Initiating Condition: LJ No Change M Difference Deviation Seismic event greater than OBE levels. Seismic event greater than OBE levels. 1) Used Alternate developer notes allowed wording since specific Control Room indication of a seismic event > OBE is not available.

Operating Mode Applicability:

Operating Mode Applicability:

All 1,2,3,4, 5, D Example Emergency Action Levels:

Emergency Action Level (EAL):

Seismic event greater than Operating Basis Earthquake (OBE) as indicated by:

a. (site-specific indication that a seismic event met or exceeded OBE limits) 1. Control Room personnel feel an actual or potential seismic event.

AND

2. The occurrence of a seismic event is confirmed in a manner deemed appropriate by the Shift Manager or Emergency Director.

Page 59 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HA5i HA5if~f~

Initiating Condition: ALERT Initiating Condition: [AS M No Change [,, Difference [F, 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 to operations, cooldown or shutdown. cooldown or shutdown. ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example 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 is warranted.

• If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no emergency classification is

1. 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 H3 AND Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability

b. Entry into the room or area is prohibited or impeded.

Reactor Building* Mode 3 and 4 Turbine Building* Mode 3

-Areas required to establish shutdown cooling AND

2. Entry into the room or area is prohibited or impeded.

Page 60 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HU3 D]

HU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change n Difference -] Deviation Hazardous Event 1) Included River water level as part of the site specific list of natural or Hazardous Event technological hazard events. The EAL values selected are the current approved Operating Mode Applicability: EALtUE values.

Operating Mode Applicability:

All 2) Changed the word "needed" to "required by Technical Specifications" in the EAL 1,2, 3,4, 5, D to be consistent with terminology used by operators and minimize confusion.

Example Emergency Action Levels: (1 or 2 or 3 or 4)

Emergency Action Level (EAL):

Note: EAL #3 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Note:

I. A tornado strike within the PROTECTED AREA.

• EAL #4 does not apply to routine traffic impediments such as fog, snow.

ice, or vehicle breakdowns or accidents.

2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.

1. Tornado strike within the PROTECTED AREA.

3. Movement of personnel within the PROTECTED AREA is impeded due to OR an offsite event involving hazardous materials (e.g., an offsite chemical

2. Internal room or area flooding of a magnitude sufficient to require manual or spill or toxic gas release). automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode.

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

3. Movement of personnel within the PROTECTED AREA is impeded due to an

5. (Site-specific list of natural or technological hazard events) offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Abnormal River level, as indicated by EITHER:

a. High river level > 594 ft.

OR

b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam #14 has failed.

Page 61 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HG7 HG7 Initiating Condition: GENERAL EMERGENCY Initiating Condition: [i No Change U Difference U Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. declaration of a General Emergency.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT Other conditions exist which in the judgment of the Emergency Director indicate substantial core degradation or melting with potential for loss of containment that events are in progress or have occurred which involve actual or IMMINENT integrity or HOSTILE ACTION that results in an actual loss of physical control of substantial core degradation or melting with potential for loss of containment the facility. Releases can be reasonably expected to exceed EPA Protective Action integrity or HOSTILE ACTION that results in an actual loss of physical control of Guideline exposure levels off-site for more than the immediate site area. the 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

NEI 99-01 Rev 6 Proposed EAL Justification HS7 HS7 m D ne Initiating Condition: SITE AREA EMERGENCY Initiating Condition: L 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 warrant declaration of a Site Area Emergency. declaration of a Site Area Emergency.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4, 5, D Example 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 indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that events are in progress or have occurred which involve actual or likely major that results in intentional damage or malicious acts; (1) toward site personnel or failures of plant functions needed for protection of the public or HOSTILE ACTION equipment 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 or access 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 not Guideline exposure levels beyond the site boundary. expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Page 63 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: ALERT Initiating Condition: [jj 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 warrant declaration of an Alert. declaration of an Alert.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2,3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential Other conditions exist which in the judgment of the Emergency Director indicate 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 that

~involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be involves probable life threatening risk to site personnel or damage to site limited to small fractions of the EPA Protective Action Guideline exposure levels, equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Page 64 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT Initiating Condition: LUh No Change [j Difference Deviation Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT. Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT.

Operating Mode Applicability:

Operating Mode Applicability:

All 1, 2, 3,4, 5, D Example 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 indicate degradation 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 facility response or monitoring are expected unless further degradation of safety systems protection has been initiated. No releases of radioactive material requiring offsite occurs. response or monitoring are expected unless further degradation of safety systems occurs.

Page 65 of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification E-HU1 E-HU1 Initiating Condition: UNUSUAL EVENT Initiating Condition: F1 No Change 1-- Difference FIDeviation Damage to a loaded cask CONFINEMENT BOUNDARY. 1) Listed 2x the site specific cask specific allowable radiation level as per Damage to a loaded cask CONFINEMENT BOUNDARY.

Certificate of Compliance No. 1014 Appendix A Section 5.7.

Operating Mode Applicability:

Operating Mode Applicability:

All 1, 2,3,4,5, D Example Emergency Action Levels:

Emergency Action Level (EAL):

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-technical specification allowable radiation level) on the surface of the spent fuel contact radiation reading:

cask.

• > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR

• > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts.

Page 66 of 66

I=xAInn N.nlmar Oijad Citiesa Annex Exelnn NucleIuar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARGI Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRfem TEDE or 5000 mRirem thyroid CDE.

Operating Mode Applicability:

1, 2,3,4,5, D Emergency Action Level (EAL):

Notes:

" The Emergency Director should declare the General _mer=.gency 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 path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

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

(1) Reading on ANY of the following radiation monitor-s greater than the reading shown for 15 miueso longer:ý (Site Specific monitor list and threshold v:alues)

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCilsec for >

15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR

2. Dose assessment using actual meteorology indicates doses at or beyond (site-speGifi dose receptor POW.) the site boundary of EITHER:

a. > 1000 mRem TEDE OR

b. > 5000 mRem CDE Thyroid OR Month 20XX QC 3-1 EP-AA-1006 (Revision XX)

r1iiesA fifimom Annmv IP=alnn Mnilanr S.'U;I l *ll rl=III 5fl kA~l.I 11 .51 l il RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS FWild- SU y...results indicate EITHER of the following at OF beyond (site specific dose

,e veptOFv, . ,, -.

Closed window do6e rates greater than 1,000 mR'~hr expected to continue for 60 minutes or longer.

Analyses of field- survey sampler, indicato thyroid CDE greater than 5,000 mrFem:

3. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates >1000 mR/hr are expected to continue for > 60 minutes.

OR

b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

The TEDE dose is set at the EPA PAG of 1000 mRrem while the 5000 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor rcadings assumes that. a release path to the enlVironment is established. if the effluent flow past an effluent monGitor is known to hav stopped due to actions to isolate the release path, then the effluent mon0itor reading is-no longer Valid. f9or clasAsi-ation purposes.

Basis Reference(s):

1. NEI 99-01 Rev 6, AG1

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-2 EP-AA-1006 (Revision XX)

(I"orl Mfiaa Annov =vm1en kimie-Igftr flu~~~~~_ý.v Fý"rffI.m.I~ vvIIiio~wz RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARS1 Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mRfem TEDE or 500 mRfem thyroid CDE.

Operating Mode Applicability:

1, 2,3, 4, 5, D Emergency Action Level (EAL):

Notes:

" The Emergency Director should declare the Site Area Emergency 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 path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

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

(1) Reading on ANY of the WoloG.ing radiation monitors greater than the reading shown fQr 15 minutes Or Ionger:

(Site spc oii aInd threshold values)

(2) Dorse assessment using actual meteorology indicates doses greater than 100 mr*em TEDE or 500 mr*eFm thyrod DE at or beyond (,ite Sp;cifc dose r*e*eptr (3) Field survey reaults indicate EITHER of the fllimneg at Or beyond (>ite EpeCific dose receptor poGint):

• Closed window dose rates greater than 100 mR'hrf expeeted to coentinlue for-60 minutes er-lengefn 9, Analyses of field survey samples indicate thyroeid CDE greater-than 500 mrwem for-one hour-of inthalation.

3. 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCllsec for > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR Month 20XX QC 3-3 EP-AA-1006 (Revision XX)

flimnrl f'ifidsa Annghv I::vy--Inn k,,rl--ghr nil an (~i*rne Annav Fva Ii~n him u,.Iaar RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

a. > 100 mRem TEDE OR

b. > 500 mRem CDE Thyroid OR

3. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates >100 mR/hr are expected to continue for > 60 minutes.

OR

b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite 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 are associated 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 and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1000 mRrem while the 500 mRfem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classific~ation based on effluent monitor readings assumes that a release path to the enVIronment is established.If the effluent filw past aI effluent onitoris known to have stopped due to actions to isolate the ree Ipah, then the effluent monAitor reading is 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, AS1 Month 20XX QC 3-4 EP-AA-1006 (Revision XX)

Exalon Nuclear Quad Cities Annex Eeo ula RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-5 EP-AA-1006 (Revision XX)

t'limarl Pifiae Annnv I=:y--nn N~t-lazar Amman r~it;~ Ann~w Fv~Inn I~J..vIoar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA1 I Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mR-em TEDE or 50 mRfem thyroid CDE.

Operating Mode Applicability:

1, 2, 3, 4, 5, D Emergency Action Level (EAL):

Notes:

• The Emergency Director should declare the Ale4-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 path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

• The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. rll for 15 minutes Or longer:.

(site sperific monitor list and threshold value&)

(2) Do.e asserssment using ac.tual meteero1rgY i.ndiates doses. greater thaRn 10 mRm..

TEDE or 50o mrem thyro-id C-DEM at or bevond (site se~ecfi dose Fece~tGr Do9intL I l *

(3) Analysis of a liquid effluent sample indieates a conc-entrantion or releas14e rate that would result in doses greater than 10 FW.mrvm TEDE Or 50 mne. thyroid MDE At or beyond (Site specific doerc Ptooint) for one hour of texposure.

(4) Field survey results indicateEITHER of the following at or beyond (cite.pecific doserecepto pint):.

" Closed window dose rats gr-eater-than 10 mR'hrexpected to continue for-60 minutes or

" Analyses of field sur.'ey samples iadiceAc thyroid CDE gr-eater thin 50 ffwem for-oef hour Of inhalatien.

Month 20XX QC 3-6 EP-AA-1006 (Revision XX)

Exalon Nuclaar Ouad Cities Annex Eeo ula RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 uCilsec for > 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

a. > 10 mRem TEDE OR

b. > 50 mRem CDE Thyroid OR

3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary

a. 10 mRem TEDE for 60 minutes of exposure OR

b. 50 mRem CDE Thyroid for 60 minutes of exposure OR

4. Field survey results at or beyond the site boundary indicate EITHER:

a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for_> 60 minutes.

OR

b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual 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 this magnitude represent an actual or potential substantial degradation of the level of safety of Month 20XX QC 3-7 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS the 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 and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more 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 mRfem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classific-ation based on offluont monitor readings assumes, that a roloase path t the *1nim e*1Int is establihe. If the AMlu, nt flow part an effluent monit is~l known to have  ;*~~.

... due stopped ';. to- ac--9tion to, igsolate-

.-rc

.. the

-.. release path, -.

then the effluen~t mon~itOF read~ing4",."9*"**"p is noe longer valid for cliiaIo proe.S Escalation of the emergency classification level would be via IC RAS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA1

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-8 EP-AA-1006 (Revision XX)

Ouad Cities Annex Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARUM Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM (site-speifir effluent releasecontr*olling document) limits for 60 minutes or longer.

Operating Mode Applicability:

1, 2, 3, 4, 5, D Emergency Action Level (EAL):

Notes:

" The Emergency Director should declare the Unusual Event event promptly upon determining that 60-iiwtesthe 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 60 minutes.

" Classification based on effluent monitor readings assumes that a release path to the environment is established. Ifthe effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

(1) Reading on ANY effluenRt radiation monitor greater than 2 times the (site specific effuen reease cOntrolling documnent) limits for 60 mninutes or longer:-

(site specific-monitor li6t and threshold values corresponding to 2 times the controlling docum~ent limits)

(2) ReadinRg on ANY e-ffluenAt radiation monFitor groAator- than 2times the alarm setpin et-ablished by a current radioactivi,' discharge permit fogr 6-0) minuItes6 Or longer.

(3) Samnple analysis for a g~asouso;r liquid release indicates a conentratieono repleamse ra-te greater than 2 times, the (site spec-ific effluenA-t release controlling document) limits-for-6-0 mninutes Or longer.

1. Reading on any of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for 2!60 minutes.

• Radwaste Effluent Monitor 1/2-1799-01 OR

" Discharge Permit specified monitor OR

2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined Control Room Panels or PPDS - Total Noble Gas Release Rate).

Month 20XX QC 3-9 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS LVACIand SGrIS Rn4dation onitor;r.1706 17 )2fond --- on Control room8 Panels-OF-OR

3. 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 a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological 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 of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

Classifiati-n based oR *oni*tGr effluent readi;ngs asumes that a Frlaso path the enyiro~nmePnt_ is established-. If the effluent- flowt pact an efflueAnt_ moitri kno~Wn to hav tpped due to actions to isolate the release path, then the effluent mon;itor reading is no longer valid foar c~lassific-atio pupoe.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

EAL #1 Basis EAL--#2-- This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.

EAL #2 Basis EAL-#-I-This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous OFrlilquid-effluent pathways.

Month 20XX QC 3-10 EP-AA-1006 (Revision XX)

(Iiind (Nfigm. Annoy l::xelnn N.PIA*r flmmnI Ci~cu Anv Fe~nn Nuclehar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS EAL #3 Basis

• AL- #3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC RAA1.

Basis Reference(s):

1. NEI 99-01 Rev 6, AU1

2. EP-EAL-0606 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-11 EP-AA-1006 (Revision XX)

I=xelon Nuclear Ouiad Cities Annex Exellon Nucleanr RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA2 I Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel.

Operating Mode Applicability:

1,2,3, 4,5, D Emergency Action Level (EAL):

=="=! 1K1Q- r3A-rU1A1AN/

'-I1 I 7 f t* *l .Iy* *I * * . . .

(2) Damage to irraidiaEtedC- fu-el reSUlting in a release of radioactivity from the fuel as inicte by 'ANY of the folloin~g radiation monitors:

(site specific liStinig Of radiation monitors, and the associated readings&,

setpoints and/40r allarm)

(3) Lowering of spent fuel pool level to (site specific Level 2 value). [320e Deve~opo Noto&J

1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

OR

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R1 Radiation Monitor reading >1000 mRemlhr Table R1 Fuel Handling Incident Radiation Monitors 0 1(2) 1705-16A Fuel Pool Rad Monitor 0 1(2) 1705-16B Fuel Pool Rad Monitor Basis:

REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.

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

CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.

Month 20XX QC 3-12 EP-AA-1006 (Revision XX)

OsiuI Cities Annex ExeIon Nuclanr RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly fuel

~ . ~a.. *"*"

~ orasnFiatýMW~

• oong assmby. ....-

. We " "--Of.-.....

Water ' .... within the levrel '-f** . spent

. . fel'" (seo poo!

D..veoe. N..tes). These events present radiological safety challenges to plant personnel 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 the plant.

This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.

Escalation of the emergency would be based on either Recognition Category A or C ICs.

EAL #1 Basis RAI *4 This EAL escalates from RAU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

EAL #2 Basis EAL-#2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, 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 a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency would be based on either Recognition Category RA-or C ICs.

Month 20XX QC 3-13 EP-AA-1006 (Revision XX)

O.iad Riesw Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Spent fuel pool Wator level at this value is Within the loWer And of the level range necessary to preVent significant dose consequencaes from direct gamrma radiation to personnel po~ferming Operations in the vicainity of the spent fuel pool. Thiscodtn refloc~te a significant losr, of spent fuel pool Water inventor; and thus iot. is aloa precursor to a loss of the ability to adequately cOOl the irradiated fuel assembles stored in the pol Escralat-ion of the Wemegency c-lassific~ation level would be via IQ; ASI1 or A82 (sag ASL2 Devoloe, ,oth).;,

Basis Reference(s):

1. NEI 99-01 Rev 6, AA2 Month 20XX QC 3-14 EP-AA-1006 (Revision XX)

n"nrl r-ifina Annav I=Y,-nn N, ielan-r

0. .~ ri (~i*igi@ Annoy Fvolnn MivIa~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARU2 Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.

Operating Mode Applicability:

1,2,3,4,5, D Emergency Action Level (EAL):

(1) a. UNPLAlNNED water level drop in the REFUELING PATHWAIAY as inrdicat by ANY of the following:

(site specific level indieatiens).

b. UNPLANNED rie inae adiation levels as indlicated by ANY of the following radiation monitors.

(site spocific list of area radiation menitor-s)

1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

" Refueling Cavity water level < 282 in. (Upper Wide Range simulated signal).

OR

• Spent Fuel Pool water level < 19 ft. above the fuel (<4 ft indicated level).

OR

• Indication or report of a drop in water level in the REFUELING PATHWAY.

AND

b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R1.

Table RI Fuel Handling Incident Radiation Monitors

• 1(2) 1705-16A Fuel Pool Rad Mon

• 1(2) 1705-16B Fuel Pool Rad Mon Month 20XX QC 3-15 EP-AA-1006 (Revision XX)

flivmA r'i#iga Anng%,w P:valnn Kii-rloor A. .~ ,I 1~.*rne Annn~v Fvair~n Mm.rIc~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.

This IC addresses a de*ease loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level deerease-loss will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available) or from any other temporarily installed monitoring instrumentation. A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance 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, AU2

2. Technical Specification 3.7.8

3. Technical Specification 3.9.6 Month 20XX QC 3-16 EP-AA-1006 (Revision XX)

flvaýA ('Mgs4m Anngav I::valnn ki~rl-,ldr RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA3 I Initiating Condition:

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

Operating Mode Applicability:

1, 2, 3, 4, 5, D Emergency Action Level (EAL):

Note:

Ifthe equipment in the listed-room or area listed in Table R3 was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted to Contr@Ioleefom G enftral Al---m Station 0 (other Site Specifi ar-eaSrooms) levels that I^\ A I I i I ll*l A ill Ip[l*

• JL--

6A)in AR WNPUNNNI-w event rosults 1n Faalialion WGRoENio Or iMDPeA RGracces t any; of the folloWing plant Fooms OF areas:

(site 6necific !irt of ...... F ilant reoms Or

v. areas with entr related Mode anpnmI*nallit

........... j V-I o

1. Dose rate geate--than> 15 mR/hr in ANY of the areas contained in Table R2:

Table R2 Areas Requiring Continuous Occupancy

" Main Control Room (Unit 1 ARM Station #22)

" Central Alarm Station - (by survey)

OR

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to any of the areas contained in Table R3:

Month 20XX QC 3-17 EP-AA-1006 (Revision XX)

flomri fifimcz Anngwv I:v,-Inn Mlill-,ar RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS Table R3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building* Mode 3 and 4 Turbine Building* Mode 3

• Areas required to establish shutdown cooling Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 to preclude or impede personnel from performing actions necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal plant proceduresmaintain norm.al plant operation, Or to perform a normF.al plant c .oldownan sh'tdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

Table R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown. This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.

For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in serviccat the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding beyond that required by procedures, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

Month 20XX QC 3-18 EP-AA-1006 (Revision XX)

Exalon Nuclear ExlnNcEar Ouad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS 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 affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

" The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

" The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

" The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

Escalation of the emergency classification level would be via Recognition Category RA, C or F ICs.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA3

2. FSAR Section 3.2 Month 20XX QC 3-19 EP-AA-1006 (Revision XX)

nmvmA ifigsim Annghw i::Ylnn N,,P-lan::r Am.~rI f~4a~ Aatui~w Fvair~n Mi urIn~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS SRU3 Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

(I) (Site specific. radiation moenitor) reading greater than (site specific value)-.

(2) Sample analysis indicates that a reactorcoolant actvity value as greater than an a.llow-able Im;it in Technical Spcfi÷atiOnc.

pe,ified

1. Offgas system radiation monitor HI-HI alarm.

OR

2. Specific coolant activity > 4.0 uClIgm Dose equivalent 1-131.

Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified 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 clad degradation should not result in the declaration of an Unusual Event.

This EAL addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity.

An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample activity (as determined by laboratory confirmation). Fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another 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, SU3

2. Technical Specifications 3.4.6

3. Technical Specifications 3.7.6 Month 20XX QC 3-20 EP-AA-1006 (Revision XX)

A. .21 (difiaa Annav I~vnlt~n k~l-I-mr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:

Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.

Operating Mode Applicability:

1,2,3 Emergency 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 I4Table 9-F-2 Month 20XX QC 3-21 EP-AA-1006 (Revision XX)

(1"nrl ('ifiaa Annav I:va,-nn K,,r-l,--r A. .~ ri t~itic~@ Annoy Fvolnn N. .rIa~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:

Loss or Potential Loss of ANY two barriers.

Operating Mode Applicability:

1,2,3 Emergency 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-3Table 9-F-2 Month 20XX QC 3-22 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FAI Initiating Condition:

ANY Loss or ANY Potential Loss of either Fuel Clad or RCS.

Operating Mode Applicability:

1,2,3 Emergency 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 more heavily 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 of Containment barrier in combination with loss or potential loss of either Fuel Cladding or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F 3Table 9-F-2 Month 20XX QC 3-23 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:

RCS Activity Operating Mode Applicability:

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

LOSS A. (Site Spcciftic ndicatiens that reactor coolant activity is grcatcr than 300uGi/gmR doc

...... *quivI 3!)Coolant activity > 300 uCilgm Dose Equivalent 1-131.

Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 ýiCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-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-9F-4Table 9-F-2 Month 20XX QC 3-24 EP-AA-1006 (Revision XX)

l=x*lon NuclAar Oijad Cities Anne~x Exelnn Nuclea~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS A-.-1. Plant conditions indicate Primary Crontainment flooding is required.

POTENTIAL LOSS A-.2. RPV water level cannot be restored and maintained abovo (site specific RPV wateF. level GOg

...... .. hE , . fuel)..,,,> -142 inches (TAF)

3. RPV water level cannot be determined.

Basis:

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

Loss 2AThreshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.

This is identified in the BWROG F-PGsEOPs/SAGsSAMGs when the phrase, "Primary Containment Flooding Is Required," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "Primary containment flooding required,"

also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.

Potential Loss 2-,AThreshold #2 and #3 Basis This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.

The RPV water level threshold is the same as RCS baF-ei-Barrier Loss threshold-24A.

Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier 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, RPV-wat&RPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the Month 20XX QC 3-25 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order 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 drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

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

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 to deliberately lower RPV water level to the top of activ" fuel in order to reduce reactor power. RPV water level is then contro.lled beteen the top of activ; fu!l and the Minimum Steam Cooling RPV V^ater Level (..S.RV*ý.- Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA5-MA3 or SS5-MS3 will dictate the need for emergency classification.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, Table9 F4,Table 9-F-2 Month 20XX QC 3-26 EP-AA-1006 (Revision XX)

MwnA Pifigsim Annmv I:vylnn kielaar Am. a ~4 r~;4i~a Anr~aw ~vciInn MmmI~ar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS A. Primary containment radiatonR MGnitor reading greater than (site specific. value)

Drywell radiation monitor reading > 6.65 E+02 R/hr.

Basis:

L-ess-4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 [tCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of 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 RCS Barrier RC5 Loss Tthreshold 44-Asince it indicates a loss of both the Fuel Clad Barrier and 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 Potential Loss threshold associated with Primary Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-9-F-4Table 9-F-2

2. Core Damage Assessment Methodology Month 20XX QC 3-27 EP-AA-1006 (Revision XX)

Exalon Nuclear Ouad Cities Annex Exe.... Nu.. e..

RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS 1A. Any condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.

POTENTIAL LOSS 2A. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.

Basis:

Loss Threshold #1 Basis L*,*&A This 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 #2 Basis Petentwal Io L.,6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table -F-3Table 9-F-2 Month 20XX QC 3-28 EP-AA-1006 (Revision XX)

Mind (Nfiaa Annoy IFyalnn khielanr flu .~ rI ('~i*i~@ Ann~w Fvalnn hi.., ia2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS

1. RPV water level cannot be restored and maintained abovo (site "pecific RPV water level correpondin't the top of active fuel)

'> -142 inches (TAF)

Gr-OR

2. RPV water level cannot be determined.

Basis:

Lgs&-2L.A RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.

The RPV water level threshold is the same as Fuel Clad bairieF Barrier FC2 Potential Loss threshold--2-A. Thus, this threshold indicates a Loss of the RCS barrier and Potential 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, RPV-wateFRPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order 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 drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The Month 20XX QC 3-29 EP-AA-1006 (Revision XX)

Exelan NUC102r Oijnd Cities Annex Exl... Nu.le.r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

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 to deliberately lower RPV water level to the top of arct'e fue-l in order to reduce reactor power. RPV water leVel i6 then controlled between the top of a.t.ve fuel and the Minimum Steam.Co

• oling RPV Water Le'vel (MSCGRV^AW.Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA5-MA3 or SS&MS3 will dictate 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-§-F-3Table 9-F-2 Month 20XX QC 3-30 EP-AA-1006 (Revision XX)

Clmiorl fifiaa Anng%,v IPv--Ign khme-l-or A. marl r~;*;~ Anv~av Fv~Lnn Mm.iIaar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

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

LOSS A. Primary containment pressure greater than (site specific"value) due to RCS leakage-.

1. Drywell pressure >2.5 psig.

AND

2. Drywell pressure rise is due to RCS leakage Basis:

The (site spe.ific.value)> 2.5 psig primary containment pressure is the 4ryweI4Drywell high pressure setpoint which indicates a LOCA by automatically initiating the-ECCS-of oguivalent makeup system.

The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.

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

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications 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---F--3Table 9-F-2 Month 20XX QC 3-31 EP-AA-1006 (Revision XX)

rlimexri ('Mga w~s~ ~

Annimv

,-~. .5 .~n Fval~nn

~

kuir ibr S S RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:

RCS Leak Rate Operating Mode Applicability:

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

LOSS Al. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC line break. in ANY of the following: (site specific. systems with potential for high energy line OR

82. Emergency RPV Depressurization is required.

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

a4. Secondary Containment area temperature > QGA 300 MaximumMax Normal Ooperating Tempe~akelevels.

OR b-2. Secondary Containment area radiation level > QGA 300 Maximum Max-Normal Ooperating Area Radiation ,Level.

Basis:

UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.

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 that the on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.

Month 20XX QC 3-32 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Loss Threshold #1 Basis-3A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the CoGntrol Room, the RCS barrier Loss threshold is met.

Loss Threshold #2 Basis-4B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the 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 the RCS is being vented into the Torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

Potential Loss Threshold- #3 Basis 3-1A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.

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

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam 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 to a Site Area Emergency when combined with Containment Barrier CT6 Loss Tthreshold

1. 13A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Basis Reference(s):

Month 20XX QC 3-33 EP-AA-1006 (Revision XX)

Ouad Citie Annex Ex.lon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION

1. NEI 99-01 Rev 6, 1b-9-FTable 9-F-2 Month 20XX QC 3-34 EP-AA-1006 (Revision XX)

flmitsri ('ifina Annnv i=Y'*h*n kJi iijhmr Au. ~ ii t~i*iao** Anna~v r~. ** 3s 1S fl

• RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:

Primary Containment radiation Operating Mode Applicability:

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

LOSS

1. Drywell radiation monitor reading > I00R/hr.

A. Primary containment radiation reading greatcr than (Site pecific value).

Basis:

Less-4-A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 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 RCS Barrier only.

There is no Potential Loss threshold associated with Primary Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-9-F-3Table 9-F-2

2. Calc. EP-EAL-0611 Month 20XX QC 3-35 EP-AA-1006 (Revision XX)

tlmmA Mfiaa Annav I::valr'*n Klir-lanr flu a~ rI (~j*jgai@

~t4a . .. Ann~v rn .. .n Fvalnn S**U *Miur6~r

• • ~c~

RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS Al. Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.

POTENTIAL LOSS A2. Any condition in the opinion of the Emergency Director that indicates Potential I Loss of the RCS Barrier.

Basis:

Loss 6-AThreshold #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 #2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-Q-F4Table 9-F-2 Month 20XX QC 3-36 EP-AA-1006 (Revision XX)

Onnrl Cities Annexw ExeIon Nucleanr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

POTENTIAL LOSS A.-Primary containment flooding is required.

Basis:

Potential Loess 2.A.

The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGsEOPs/SAGsSAMGs specify the conditions that require primary containment flooding. When primary containment flooding is required, the EPGsEOPs are exited and SAGsSAMGs are entered. Entry into SAGsSAMGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt 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 a General Emergency.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table-9-F4Table 9-F-2 Month 20XX QC 3-37 EP-AA-1006 (Revision XX)

Ouad Cities Annex Exelon Nuclear I=x*lon Nucilear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Primary Containment Conditions Operating Mode Applicability:

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

LOSS Al. UNPLANNED rapid drop in prmry .tainmentDrywell pressure following P.a GentaimentDrywell pressure rise.

OR B2. Prim.a.y contai.me nDrywell pressure response not consistent with LOCAI conditions.

POTENTIAL LOSS A3. Prim.ay contai÷nmntDrywell pressure greater than (site specific Value)> 56 psig and rising.

OR

84. (site s eXPlosive mixture) exists inside prima.y containment a. Drywell or

.peci..

torus hydrogen concentration > 6%.

AND

b. Drywell or torus oxygen concentration > 5%.

OR G5. HTLG-Heat Capacity Limit (QGA 200, Figure M)exceeded.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Loss !-A-a n-d-,BThreshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to dPywellDrywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity. Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. 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 the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment Month 20XX QC 3-38 EP-AA-1006 (Revision XX)

f-l"nel Mae Annoy Pgi:n~rn Nh:PlAd-ar A. .~ .1 (~i$i~o Annoy Fvolnn ItJ,,rIo~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.

Potential Loss 4-tAThreshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.

Potential Loss 4.BThreshold #4 Basis If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.

Potential Loss 4-.GThreshold #5 Basis The HCat Capacity Temperatufe Limit (HpTr) is the highest Torus twmperature from Eergency RP2V D)Pep-ressu-ri-atinn will no-t raais-e:-

which equipmcnt within the Torus which may be required to operate whenM t~he RPV is-

• Torus pressure aboeve Priimar,' Containment PreSSUre Limit A, whilec the rate of energy transfer from the RP2V to the containment is greater than the capacityo the GentainmenRt vent.

The HCTL is a function of RPV pressure, Torus temperature and 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 and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

Basis Reference(s):

1. NEI 99-0 1 Rev 6, Table 9 I=3Table 9-F7-2 Month 20XX QC 3-39 EP-AA-1006 (Revision XX)

r1minel Pifigaa Annnv IPv,-Inn M, iP-lar Am, ~ rI(*iniaAnnn~v FvaIr~n 1Ji.rIa~r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

POTENTIAL LOSS A.PriMaR' containmel# radiation monGitor reading greatker than (site Spocvific value)

1. Drywell radiation monitor reading > 1.55 E+03 R/hr.

Basis:

There is no Loss threshold associated with Primary Containment Radiation.

Poten~tial Loss 4. A.

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. 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 NuclearPower Plant 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 exist- there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential 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-3Table 9-F-2

2. Core Damage Assessment Methodology Month 20XX QC 3-40 EP-AA-1006 (Revision XX)

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t* Annaw St**U fl Fvc S** innS UNh.rIc~r

.1, RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:

Primary Containment Isolation Failure Operating Mode Applicability:

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

LOSS Al. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR B2. Intentional Pprimary CGontainment venting/purging per EOP's or SAMGs due to accident conditions.

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

1. Secondary Containment area temperature > QGA 300, MaximumMa* Safe Qoperating TenpeFakwelevels.

OR

2. Secondary Containment area radiation level > QGA 300, MaximumMa* Safe Ooperating Radiatio* I-levels.

Basis:

UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.

These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss 3-.AThreshold #1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument 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 breached and thereby creates a significant pathway to the environment. Examples include unisolable Main Steamline, HPCI or RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.

Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Month 20XX QC 3-41 EP-AA-1006 (Revision XX)

Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. 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 but should be evaluated using the Recognition Category A-R ICs.

Loss 3,BThreshold #2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the dPywelDrywell high pressure scram setpoint) does not meet the threshold condition.

Loss 3.CThreshold #3 Basis The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each 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 the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam 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 Lioss &AThreshold #3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Basis Reference(s):

Month 20XX QC 3-42 EP-AA-1006 (Revision XX)

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-~ 5. i=yalnn N"Aal*nr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION 1 NEI 99-01 Rev 6, Table-Q-FTable 9-F-2 Month 20XX QC 3-43 EP-AA-1006 (Revision XX)

flimri Pifigha Annghv l~v--lnn klmrI-,ar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS Al. Any condition in the opinion of the Emergency Director that indicates Loss of the I Containment Barrier.

POTENTIAL LOSS A2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.

Basis:

Loss $-AThreshold #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 #2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table9 F- 3Table 9-F-2 Month 20XX QC 3-44 EP-AA-1006 (Revision XX)

rlimýri ('Mga Anngv Fwi:vlnn NllrAaar Au ~rI r~;*;~ Ann~v Fv~Inn Mmmr.6~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSG1 Initiating Condition:

Prolonged loss of all Off-site and all On-Site AC power to emergency busses.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the General Emerge..yevent promptly upon determining that (site specific hOurS) the applicable time has been exceeded, or will likely be exceeded.

1---a. Loss of ALL offsite and- ALL, on'it- AC power to ("it Specific em.ergen.y buses)unit ECCS busses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND 3b. EITHER of the following:

a. Restoration of at least one em&§enGy-unit ECCS bus in < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is notles.

than speGift, heus-i

.(site . 'net likely.

OR

b. RPV water level cannot be restored and maintained > -166 inches.

(Site specific indic-ation o-fan inability to adequately remRove heat fromn the core)

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat Month 20XX QC 3-45 EP-AA-1006 (Revision XX)

rl"nrl (ifiaa Annov P~valnn khnnllanr flu .~ r4 (~i*iao Ann~v Fv~Inn Mumr1~2r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of -- o Or mo -- ny fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event 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 not be used as a basis for delaying a classification upgrade. The goal is to maximize the time 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 power results in parameters that indicate an inability to adequately remove decay heat from the core.

Basis Reference(s):

1. NEI 99-01 Rev 6, SG1

2. UFSAR 8.3 Month 20XX QC 3-46 EP-AA-1006 (Revision XX)

Dinid Citiest Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY M$SI Initiating Condition:

Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Sito Area r*,-..genc-yevent promptly upon determining that the applicable time 165- inues-has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite and ALL -n'itoA*C Power to (site specific emnergeny bus.s)unit ECCS busses fQo 15 minutesorO lonr.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RAG1, FG1 or MSGI.

Month 20XX QC 3-47 EP-AA-1006 (Revision XX)

Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis Reference(s):

1. NEI 99-01 Rev 6, SS1

2. UFSAR 8.3 Month 20XX QC 3-48 EP-AA-1006 (Revision XX)

IExAlnn Nuclanr 0"nd Cifian Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSA1 Initiating Condition:

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

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the eventAled promptly upon determining that the applicable time 4§5rFinutes-has been exceeded, or will likely be exceeded.

1. AC power capability to unit ECCS busses reduced to only one of the following power sources for > 15 minutes.

• Reserve auxiliary Transformer TR-12 (TR-22)

• Unit Auxiliary Transformer TR-1 1 (TR-21)

• Unit Emergency Diesel Generator

• Shared Emergency Diesel Generator

• Station Blackout Diesel Generator

• Unit crosstie breakers

a. AC PoWer capability to (site specific e*ne*geRnY buses) is re-dued, to a single power Source forF 15 minutes or longer.

AND 2b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC MSU1.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

Month 20XX QC 3-49 EP-AA-1006 (Revision XX)

Mind Civinv An.n. Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY

" A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

" A loss of all offeite power and- lo9s86 Of a~ll emer9gency power soreFes (o.g., GAit die-Sel generators)-' wih snle train of emergencay buses, being back fed from th Unit mai gnrtor.

" A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Escalation of the emergency classification level would be via IC MSS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA1

2. UFSAR 8.3 Month 20XX QC 3-50 EP-AA-1006 (Revision XX)

ula Exelon Nuclear Eeo Ouad Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Msul Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unu'ual E,-cntevent promptly upon determining that the applicable time 4Min#ute&s-has been exceeded, or will likely be exceeded.

-.Loss of ALL offsite AC power capability to unit ECCS busses (site-speGifiG

.....gen.y bue)

.. 9F for > 15 minutes-er- i.

Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders 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 power source(s) is available to the emergency buses, whether or not the buses are powered from it. (e.g. unit cross-tie breakers)

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC MSAI.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU1

2. UFSAR 8.3.1 Month 20XX QC 3-51 EP-AA-1006 (Revision XX)

Explon Nuclear Ouad Cities Annex Ex......uc...r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSG21 Initiating Condition:

Loss of all AC and Vital DC power sources for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the General *,e*,rgenr. vent promptly upon determining that the applicable time 4-5miRutes-has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to unit ECCS busses.

AND

2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to vital busses.

AND

3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.

AND

4. All AC and Vital DC power sources have been lost for >15 minutes.

buses) for 1 -5 minutesor lOFnger.

AND h Ipdinat.~edr "r'ltin i6 lrwti than (Rite rnne~ific W6i "nltinnqV AAAIII

• l*V J nnl A'*hhI yl I K (cite munific Vitl DvC hicerqw fio 15 min,udte or long.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

Month 20XX QC 3-52 EP-AA-1006 (Revision XX)

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%fA -aI. naw FvlnM.rI RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both-all EALs threshelds-are met.

Basis Reference(s):

1. NEI 99-01 Rev 6, SG8

2. UFSAR 8.3.2

3. Technical Specifications B.3.8.4 Month 20XX QC 3-53 EP-AA-1006 (Revision XX)

fluisd Cities Annex Exelon Nuclea~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month 20XX QC 3-54 EP-AA-1006 (Revision XX)

fliinrl fMae Anngi-tr =vgmlrn Klimr-lAmor Al I ~tI i4la~ Aijnav ~vaIr~n MumIn~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSS28 Initiating Condition:

Loss of all vital DC power for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Site Area Emergency-vent promptly upon determining that the applicable time 41-minutes-has been exceeded, or will likely be exceeded.

kIdi*ated-vVoltage is < 105 VDC less than (site specific. bus voltage value) on 125 VDC battery busses #1 and #2ALL (site specific V--ital D, busses) for >15 minutes OF.4Gnge

'Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a loss of Vital DC power which compromises the ability to monitor and 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 power losses.

Escalation of the emergency classification level would be via ICs RAG1, FG1 or MSG28.

'Basis Reference(s):

1. NEI 99-01 Rev 6, SS8

2. UFSAR 8.3.2.1.1

3. Technical Specifications B3.8.4 Month 20XX QC 3-55 EP-AA-1006 (Revision XX)

fla.iesI (fifin ý185Rýý Anniwv I:=v--Itn Nimil--nmr

%ftý RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSS35 Initiating Condition:

Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

2. All manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.

AND

3. EITHER of the following conditions exist:

• RPV water level cannot be restored and maintained > -166 inches OR

" Heat Capacity Limit (QGA 200, Figure M) exceeded.

(Site spec-ific- ind-ic-ation of an inability to adequately remov.e heat fromn the cree)

(Site 6pecifi indication of an inability to adequately remo-ve he-at frm.-. t~he RC;S)

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection,actins, to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

Month 20XX QC 3-56 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.

Escalation of the emergency classification level would be via IC RAG1 or FG1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SS5 Month 20XX QC 3-57 EP-AA-1006 (Revision XX)

ExAIon Nuclear Ousari ities Annex Exelon Nucla~r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSA36 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

1. An-aAutomatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

2. Manual / ARI actions taken at the reactr control cnc.s.clReactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged Month 20XX QC 3-58 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MSS35. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC MSS35 or FS1, an Alert 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 are included 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, SA5 Month 20XX QC 3-59 EP-AA-1006 (Revision XX)

O.,ad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSU31 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

1. a. Ap,-aAutomatic scram -did not shutdown the reactor as indicated by Reactor Power > 5%.

.AND

b. A-sSubsequent manual action taken at the reactor control consl Reactor Console is successful in shutting down the reactor.

OR

2. a. A-mManual scram / ARI did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

b. EITHER of the following:

1. A-sSubsequent manual / ARI action taken at the reactor control GeRseleeReactor Console is successful in shutting down the reactor.

OR

2. A-sSubsequent automatic scram / ARI -is successful in shutting down the reactor.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in 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 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

Month 20XX QC 3-60 EP-AA-1006 (Revision XX)

I=xelon Nuclear Q~mzrl iCigti Annex Exellon Nucleanr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor tscram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MSA35. Depending upon the plant response, escalation is also possible via IC FAI. 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 setpoint testing), the following classification guidance should be applied.

" If the signal generated as a result of plant work causes a plant transient that created a real condition that should have included an automatic reactor scram and the RPS fails 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 but should 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 EALs are not applicable and no classification is warranted.

Month 20XX QC 3-61 EP-AA-1006 (Revision XX)

Exelon Nuclear Quad Cities Annex Eeo ula RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis Reference(s):

1. NEI 99-01 Rev 6, SU5 Month 20XX QC 3-62 EP-AA-1006 (Revision XX)

QU2d Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSA42 Initiating Condition:

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

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the eventAle~ promptly upon determining that the applicable time 45 minutes-has been exceeded, or will likely be exceeded.

1. a. AR,-UNPLANNED event results in the inability to monitor one Or mRcANY Table Mlef the*f*llowing parameters from within the Control Room for > 15 minutes oF loRge-.

[see table below]

JB WR parameter&41t Reactor PoWer Table M1 Control Room Parameters

• Reactor Power R"V Water-Level

• RPV Water Level

• RPV Pressure "VX PrcSSUrc

• Drywell Pressure Primary Containment

• Torus Level

• Torus Temperature Suppression Pool Level SUPPrcssion Pool Temperatur-e AND

b. Any Table M2ef the, fll*wi* g transient events in progress.

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" EGGS (SI) actuation Month 20XX QC 3-63 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Table M2 Significant Transients

" Turbine Trip

" Reactor Scram

• ECCS Activation

" Recirc. Runback > 25% Reactor Power Change

" Thermal Power oscillations > 10% Reactor Power Change Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product 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 er mereany of I the listed 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). For example, the reactor power level cannot be determined from any analog, computer point, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly 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 action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV-leve!RPV water level and RCS heat removal.

Month 20XX QC 3-64 EP-AA-1006 (Revision XX)

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RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY The loss of the ability to determine neo Or mereany of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for ene or-moreany of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted 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 plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RAS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA2 Month 20XX QC 3-65 EP-AA-1006 (Revision XX)

eliinrf r-ifina Annav I=walnn M, r~lakr A ui~rI (~i*iaa Ann~v FvMnn Ma.,,Igimr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY MSU421 Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL):

Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 15 minutes-has been exceeded, or will likely be exceeded.

a- An-UNPLANNED event results in the inability to monitor one orFmoRe.ANY Table M1 parameters 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 the fglloWin. Dq...ar.A.t rs, from. Within the Control Room. for 15 m;inuite

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4. Reactor Power- 6. Reactor. Pow 57--

8.RVWater- L" &RCS

9. Leve1 10-.RPX' Pressur-e 11. -R-CSFP-r-essur e

12. Prniw 12. In Corv./Come Containment Exit Temp,,rt,-

14. uppessin P15. Levels in at least Leve (site-speetie

17. Steam Cencrator Temper-ataf AffliliniffOf

___ Wate-FNOW Month 20XX QC 3-66 EP-AA-1006 (Revision XX)

ula Exallon Nuclear Eeo Ouad Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 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 plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically 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 Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one OF moreany of the listed 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). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated 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 it significantly 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 action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine eoe or mereany of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one er mereany of the listed parameters are lost, then the ability to determine 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 the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC MSA42-.

Month 20XX QC 3-67 EP-AA-1006 (Revision XX)

rl"nrl ifine Annav FPylnn umlrlnr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Basis Reference(s):

1. NEI 99-01 Rev 6, SU2 Month 20XX QC 3-68 EP-AA-1006 (Revision XX)