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Attachment 11: Discussion of Revision to the Radiological Emergency Plan Annex for Three Mile Island Nuclear Station, Cover Through Page TMI 3-79
	ML14164A079
Person / Time
Site:	Three Mile Island
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-1009, NEI 99-01, Rev. 6
	Download: ML14164A079 (146)
	v • d • e

ATTACHMENT 11 DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FOR THREE MILE ISLAND NUCLEAR STATION EP-AA-1 009 Enclosures 0 Enclosure 11 A - EAL Comparison Matrix Document 0 Enclosure 11B - EAL Red-Line Basis Document 0 Enclosure 11C - EAL Basis Document

NEI 99-01 REVISION 6 DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS ATTACHMENT 11 DISCUSSION OF REVISION TO THE RADIOLOGICAL EMERGENCY PLAN ANNEX FOR THREE MILE ISLAND STATION Exelon Generation.

NEI 99-01 Rev 6 [Proposed EAL Justification Initiating Condition - GENERAL EMERGENCY AG1 Initiating Condition:

RG1 E No Change f Difference ElDeviation 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 "Classification based on effluent monitor readings assumes that a 1,2,3.4,5,6,D release path to the environment is established." To the third bullet in order to Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Levels (EAL): delete the following from the basis section " Classification based on effluent monitor readings assumes that a release path to the environment is established. If Notes: Notes: the effluent flow past an effluent monitor is known to have stopped due to actions

The Emergency Director should declare the General Emergency promptly The Emergency Director should declare the event promptly upon determining to isolate the release path, then the effluent monitor reading is no longer valid for upon determining that the applicable time has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded. classification purposes." This allows for more timely classification since all the exceeded. basis information pertaining to Note bullet 3 will be contained in the IC and therefor If an ongoing release is detected and the release start time is unknown, readily available on the 11x17 procedure matrix used by the SM.

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.

Classification based on effluent monitor readings assumes that a release path

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

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

1. Reading on any of the following radiation monitors greater than the 1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 reading shown for 15 minutes or longer: minutes.

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

b. > 5000 mRem CDE Thyroid OR

3. Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point): 3. Field survey results at or beyond the site boundary indicate EITHER:

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

Analysis of field survey samples indicate thyroid CDE greater than OR 5000 mrem for one hour of inhalation. b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

Table Ri Effluent Monitor Thresholds Effluent Monitor General Emergency RM-G-25 (Cond Offgas) 9.53 E+06 mR/hr RM-A-8GH (Station Vent) 3.09 E+05 cpm RM-G-24 (RB Purge) 5.55 E+05 mRPhr RM-A-14 (ESF Vent) 6.66 E+02 uCi/cc Page 1 of 66

NEI 99-01 Rev 6 Proposed EAL j Justification AS1 RS1 Initiating Condition - SITE AREA EMERGENCY Initiating Condition: D No Change E 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:

1,2,3,4,5,6,D 2) Added "Classification based on effluent monitor readings assumes that a Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Levels (EAL): release path to the environment is established." To the third bullet in order to delete the following from the basis section "Classification based on effluent Notes: Notes: monitor readings assumes that a release path to the environment is established. If

The Emergency Director should declare the General Emergency promptly

The Emergency Director should declare the event promptly upon determining the effluent flow past an effluent monitor is known to have stopped due to actions upon deterrining that the applicable time has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded. to isolate the release path, then the effluent monitor reading is no longer valid for exceeded. classification purposes." This allows for more timely classification since all the

If an ongoing release is detected and the release start time is unknown, basis information pertaining to Note bullet 3 will be contained in the IC and therefor

" if an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes. readily available on the 1 1x17 procedure matrix used by the SM.

assume that the release duration has exceeded 15 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 due to to the environment is established. If the effluent flow past an effluent monitor is actions to isolate the release path, then the effluent monitor reading is no known to have stopped due to actions to isolate the release path, then the longer valid for classification purposes. effluent monitor reading is no longer valid for classification purposes.

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

1. Reading on any of the following radiation monitors greater than the 1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 reading shown for 15 minutes or longer: minutes.

OR (site specific monitor list and threshold values) 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:

a. >100mRemTEDE

2. Dose assessment actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose OR receptor point)

b. > 500 mRem CDE Thyroid OR

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

specific dose receptor point):

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

Closed window dose rates greater than 100 mR/hr expected to continue for > 60 minutes.

continue for 60 minutes or longer.

OR I

Analysis of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation. b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.

Table R1 Effluent Monitor Thresholds Effluent Monitor Site Area Emergency RM-G-25 (Cond Offgas) 9.53 E+05 mR/hr RM-A-8GH (Station Vent) 3.09 E+04 cpm RM-G-24 (RB Purge) 5.55 E+04 mR/hr RM-A-14 (ESF Vent) 6.66 E+01 uCi/cc Page 2 of 66

NEI 99-01 Rev 6 ] Proposed EAL ] Justification AA1 RA1 Initiating Condition - ALERT Initiating Condition: H:1 No Change E Difference -I- 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 "Classification based on effluent monitor readings assumes that a release path to the environment is established." To the third bullet in order to delete the 1, 2, 3,4, 5, 6, D following from the basis section "Classification based on effluent monitor readings Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Levels (EAL): 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 Note: Note: release path, then the effluent monitor reading is no longer valid for classification

The Emergency Director should declare the event promptly upon purposes." This allows for more timely classification since all the basis information

The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be pertaining to Note bullet 3 will be contained in the IC and therefor readily available determining that the applicable time has been exceeded, or will likely be exceeded. on the 11x17 procedure matrix used by the SM.

exceeded.

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

Ifan ongoing release is detected and the release start time is unknown, assume 3) Calculations were performed, in accordance with (lA) guidance provided in NEI assume that the release duration has exceeded 15 minutes. that the release duration has exceeded 15 minutes. 99-01 revision 6 EAL AA1, to determine the effluent monitor response for a

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

Classification based on effluent monitor readings assumes that a release radioactive liquid release and a WGDT release via the normal site release pathway.

to actions to isolate the release path, then the effluent monitor reading is path to the environment is established. If the effluent flow past an effluent The release would contain activity equivalent to provide 10mrem TEDE or 50mrem monitor is known to have stopped due to actions to isolate the release path, then thyroid CDE at the site boundary. The calculation determined the effluent monitor no longer valid for classification purposes. the effluent monitor reading is no longer valid for classification purposes. responses would be >110% of the instruments maximum range and as such, lAW

The pre-calculated effluent monitor values presented in EAL #1 should be

The pre-calculated effluent monitor values presented in EAL #1 should be used for NEI 99-01 Rev 6 guidance, was not included in this EAL.

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

1. Readings on ANY Table R1 Effluent Monitor > Table RI value for > 15 minutes.

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

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

(site-specific monitor list and threshold values) a. > 10 mRem TEDE OR

b. > 60 mRem CDE Thyroid

2. Dose assessment actual meteorology indicates doses greater than 10 OR mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose 3. Analysis of a liquid effluent sample indicates a concentration or release rate that receptor point) would result in doses greater than EITHER of the following at or beyond the site boundary

3. Analysis of'a liquid effluent sample indicates a concentration or a. 10 mRem TEDE for 60 minutes of exposure release rate that would result in doses greater than 10 mrem TEDE OR or 50 mrem thyroid CDE at or beyond (site-specific dose receptor b. 60 mRem CDE Thyroid for 60 minutes of exposure point) for one hour of exposure. OR

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

4. Field survey results indicate EITHER of the following at or beyond (site a. Gamma (closed window) dose rates > 10 mR/hr are expected to specific dose receptor point): continue for >360 minutes.

OR

Closed window dose rates greater than 10 mR/hr expected to

b. Analyses of field survey samples indicate a 60 mRem CDE continue for 60 minutes or longer.

Thyroid for 60 minutes of inhalation.

Analysis of field survey samples indicate thyroid CDE greater than Table R1 Effluent Monitor Thresholds 50 mrem for one hour of inhalation. Effluent Monitor Alert RM-G-25 (Cond Offgas) 9.53 E+04 mR/hr RM-A-8GH (Station Vent) 3.09 E+03 cpm RM-G-24 (RB Purge) 5.55 E+03 mR/hr RM-A-14 (ESF Vent) 6.66 E+00 uCi/cc Page 3 of 66

0 NEI 99-01 Rev 6 Proposed EAL IJustification Initiating Condition - UNUSUAL EVENT AU1 Initiating Condition:

RU1 1- No Change M Difference I 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.

2) Added "Classification based on effluent monitor readings assumes that a release Operating Mode Applicability: All Operating Mode Applicability:

path to the environment is established." To the third bullet in order to delete the 1, 2, 3,4,5, 6, D following from the basis section "Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Levels (EAL):

past an effluent monitor is known to have stopped due to actions to isolate the Note: Note: release path, then the effluent monitor reading is no longer valid for classification purposes." This allows for more timely classification since all the basis information

The Emergency Director should declare the Alert promptly upon

The Emergency Director should declare the event promptly upon pertaining to Note bullet 3 will be contained in the IC and therefor readily available determining that 60 minutes has been exceeded, or will likely be determining that the applicable time has been exceeded, or will likely be on the 1 1x17 procedure matrix used by the SM.

exceeded. exceeded.

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

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

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

Classification based on effluent monitor readings assumes that a release indicating that the release path is isolated, the effluent monitor reading is path to the environment is established. If the effluent flow past an effluent no longer valid for classification purposes. 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 effluent radiation monitor greater than 2 times the (site-specific effluent release controlling document) limits for"60 minutes or longer: 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit (site-specific monitor list and threshold values corresponding to 2 times for a 60 minutes.

the controlling document limits)

RM-L-6, Radwaste Discharge

2. Reading on ANY effluent radiation monitor greater than 2 times the

RM-L-12, IWTS I IWFS Discharge alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

RM-A-7, Waste Gas Decay Tank Discharge

Discharge Permit specified monitor

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. 2. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 60 minutes.

OR

3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of Z 60 minutes.

Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Evet RM-G-25 (Cond Offgas) 1.09 E+03 mRlhr RM-A-8GH (Station Vent) 7.03 E+01 cpm RM-G-24 (RB Purge) 6.34 E+01 mR/hr RM-A-14 (ESF Vent) 7.60 E-02 uCi/cc Page 4 of 66

NEI 99-01 Rev 6 Proposed EAL Justification AG2 RG2 Initiating Condition -- GENERAL EMERGENCY D No Change E Difference 1 Deviation 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 Operating Mode Applicability: All EA-1 2-051, it is recommended that this EAL be implemented when the enhanced spent fuel pool level instrumentation is available for use. The completion of the Example Emergency Action Levels: enhanced SFP level indicators and need for the inclusion of this EAL is being tracked in accordance with Exelon Generation Company, LLC's Initial Status NOTES: The Emergency Director should declare the General Emergency Report to March 12, 2012 Commission Order Modifying Licenses with Regard for promptly upon determining that 60 minutes has been exceeded, Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated or will likely be exceeded October 25,2012.

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 1 Proposed EAL I Justification AS2 RS2 Initiating Condition -SITE AREA EMERGENCY D No Change FI- Difference 1 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-1 2-051) dated description) October 25,2012.

Page 6 of 66

NEI 99-01 Rev 6 1Proposed EAL I Justification AAZ Initiating Condition - ALERT Significant towering of water level above, or damage to, irradiated fuel.

Initiating Condition:

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

Operating Mode Applicability: All Operating Mode Applicability:

2) EAL #3 not used in accordance with the discussion in Section 1.4, NRC Order 1, 2, 3, 4, 5, 6,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 Levels (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

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel fuel as indicated by ANY of the following radiation monitors: October 25,2012.

as indicated by ANY Table R1 Radiation Monitor reading >1000 mRem/hr (nite-specific listing of radiation monitors, and the associated readings, setpoints and/or alarms)

Table R2 Radiation Monitors

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

RMS Area Monitored Mode ALL RM-G-9 FHB Bridge Rad Monitor 5,6 RM-G-6 RB Auxiliary Bridge Rad Monitor 5,6 RM-G-7 RB Main Bridge Rad Monitor Page 7 of 66

NEI 99-01 Rev 6 = Proposed EAL I Justification AU2 RU2 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change 1:1 Difference F 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, 6, D Example Emergency Action Levels: Emergency Action Levels (EAL):

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

AND (site-specific level indications). b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R2.

AND Table R2

b. UNPLANNED rise in area radiation levels as indicated by ANY of the Radiation Monitors following radiation monitors.

RMS Area Monitored Mode ALL (site-specific list of area radiation monitors) RM-G-9 FHB Bridge Rad Monitor 5,6 RM-G-6 RB Auxiliary Bridge Rad Monitor 5,6 RM-G-7 RB Main Bridge Rad Monitor Page 8 of 66

NEI 99-01 Rev 6 Proposed EAL I Justification AA3 RA3 Initiating Condition - ALERT Initiating Condition: E No Change 1 Difference Deviation Radiation levels that impede access to equipment necessary for normal plant Radiation levels that impede access to equipment necessary for normal plant 1) Listed site specific plant rooms and areas with identified mode applicability to operations, cooldown or shutdown. operations, cooldown or shutdown. ensure timely classification.

Operating Mode Applicability: All Operating Mode Applicability:

1,2, 3, 4, 5. 6,D Example Emergency Action Levels: (1 or 2) Emergency Action Levels (EAL):

Note: If the equipment in the listed room or area was already inoperable, or out Note: If the equipment in the room or area listed in Table R4 was already inoperable, of service, before the event occurred, then no emergency classification is or out of service, before the event occurred, then no emergency classification is warranted warranted 1 Dose rite greater than 15 mR/hr in ANY of the following areas: 1. Dose rate greater than 15 mR/hr in ANY of the areas contained in Table R3:

Control Room

Central Alarm Station

(other site-specific areas/rooms)

2. An UNPLANNED event results in radiation levels that prevent or significantly impede access to any of the following plant rooms or areas:

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

OR

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to any of the following Table R4 plant rooms or areas:

Table R4 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building* Modes 4, 5, and 6 Intermediate Building* Modes 4, 5, and 6 Auxiliary Building* Modes 4, 5, and 6 Fuel Handling Building* Modes 4, 5, and 6

Areas required to establish shutdown cooling Page 9 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT SU3 RU3 Initiating Condition: -- No Change M Difference F Deviation Reactor coolant activity greater than Technical Specification allowable limits. Reactor coolant activity greater than Technical Specification allowable limits. 1) Listed site-specific monitor and threshold value to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability: 2) Listed this system category EAL in the radiological category EAL section to Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 maintain consistency with current and previous revisions of Exelon EALs. This will ensure a timely classification since the threshold values are more aligned with the Example Emergency Action Levels: Emergency Action Levels (EAL): radiological category vice system category.

1. (Site-specific radiation monitor) reading greater than (site-specific value). 1. Letdown Monitor RM-L-1 alert alanm (high or low channel).

OR OR

2. Sample analysis indicates that a reactor coolant activity value is greater than 2. Sample analysis indicates that:

an allowable limit specified in Technical Specifications.

a. Dose Equivalent 1-131 specific coolant activity > 60.0 uCilgm.

OR

b. Dose Equivalent XE-1 33 specific coolant activity

> 797.0 uCi/gm.

Page 10 of 66

NEI 99-01 rev 6 Fission Product Barrier Matrix FGt Loss of any two barriers AND Loss or Potential Loss of third bamer. 1.2,3,4 FS1 Loss or Potential Loss of ANY two bamers. 1.2.3.4 FAl ANY Loss or ANY Potential Loss of either Fuel Clad or RCS 1.2.3,4 FC - Fuel C lad RC - Reactor Coolant System CT - Containment Lub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A. Operation of a standby charging (makeup) pump is required by EITHER of A. An automatic or manual ECCS (SI) the following:

Aactuation is required by EITHER of the a UNISOLABLE RCS leakage A. RCSlreactor vessel level less OR

1. RCS or SG None than (site-specific level). following: b. SG tube leakage. A leaking or RUPTURED SG is None Tube Leakage a. UNISOLABLE RCS leakage OR FAULTED outside of containment.

OR B. RCS cooldown rate greater than

b. SG tube RUPTURE. (site specific pressurized thermal shock criteriallimits defined by site specific Indications)

A. Core exit thermocouple readings greater than (site specific A. 1. (site specific criteria for entry into A. Core exit thermocouple temperature value) A. Inadequate RCS heat removal core cooling restoration procedure)

2. Inadequate readings greater than (site OR None capability via steam generatore as None AND Heat Removal specific temperature value) B. Inadequate RCS heat removal indicated by (site specific Indications). 2. Functional Restoration Procedures capability via steam generators not effective in <15 minutes.

as indicated by (site specific indications).

A, Containment radiation monitor readinggreater than (site specific 3 RCS Activity/ value) A. Containment radiation monitor A. Containment radiation monitor reading Containment OR None reading greater than (site specific value) None None greater than (she specific value)

Radiation B, (Site specific indications that reactor coolant activityis greater than 300 uCi/gm dose equivalent 1-131).

A.Containment isolation is required AND A. Containment pressure greater than (site EITHER of the following: specific value)

1. Containment integrity has been OR lost based on Emergency Director B. Explosive mixture exists inside judgement. containment.

4.Containment OR OR Integrity or None None None None Bypass 2. UNISOLABLE pathway from C. 1. Containment pressure greater than containment to the environment (site specific value) exists. AND OR 2. Less than one fulltrain of (site B Indication of RCS leakage outside of specific equipment) is operating containment per design for _15 minutes.

A,AnyCondition in the opinion of the A. Anycondition in the opinion ofthe A.ANYCondition in the opinion ot the A.Any Condition in the opieonof the A.AnyCondition in the opinion ofthe A. AnyConditon in the opinion of the Emergency

5. Emergency n Emergency Director that indicates Loss mergecy Directorthat indicates Potential Emergeecy Director tharindicates Loseat Emergency Directortsar indicates Potential Emergency Directortsar idicates Loss ot the Director that indicates Potential Loss of the Director Judgment of the Fuel Clad Barier. Loss of the Fuel Clad Barner. the RCS Barmer. Loss of the RCS Bamer. Containment Barrier. Containment Barrer.

Page II of 66

Proposed Fission Product Barrier Matdx FG1 Loss of any Iwo banriors ANDLoss or Potential Loss of third barrier. FBI Loss or Potential Loss of ANY two barrers. [fE [ FAl ANY Loss or ANY Potential Loss of either Fuel Clad or RCS OEMi~i FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category, Loss Potential Loss Loss Potential Loss Loss Potential Loss

2. UNISOLABLE RCS leakage >159gpm.

1. RCITS hot leg instruments indicate 1. Automatic or manual ESAS adtuation is OR 0 inches after lowering trend, required by EITHER of the following:

AND a. UNISOLABLE RCS leakage 3. HPI-PDRV Cooling in elfect. l e >

b LRCSor Sg None 2. In-core thermocouples are unavailable. OR OR 2. SG is FAULTED outside of containment. None Tube Leakage AND b. Steam Generator tube 4. a. RCS Pressure a 2450 psig.

3. ALL RCP's are secured. RUPTURE.

AND

b. RCS Pressure not lowering

1. T_ a _ 1800iF.

2. > 25"F Superheat AND

2. Inadequate 1. T > 14001F OR None None 2. EOP Restoration procedures not Heat Removal 3. HPF-PORV Cooling in effect. HPI-PORV Cooling in effect. effective in 1e<minutes.

1. Containment radiation monitor (RM-G-22 or RM-G-23) reading > 1.95E+03

3. Containment R/hr. Containment radiation monitor (RM-G-22 or Containment radiation monitor (RM*G-22 or Radiation / RCS OR None RM-G-23) reading > 25 Rlhr. None None RM-G-23) reading > 4.40E+03 R/hr o Adivity 2. Coolant activity > 300uCi/gm Dose Equivalent 1-131

3. Reactor Building Pressure > 55 psig

1. Containment isolation is required and and rising.

EITHER of the following: OR

a. UNPLANNED lowering in 4. Hydrogen Concentration in containment pressure or rise in Containment a 4%.

radiation monitor readings OR outside of containment in the 5. a. Reactor Budding pressure Emergency Directors judgment 30 psig indicate a loss of containment

4. Containment None None None None integrity. AND Integnty or Bypass OR b. Reactor Building Emergency

b. UNIODLABLE pathway frow cooling is less than ANY one of the containment to the f cnios environment exists.

OR SPRAY COOLERS

2. Indication of RCS leakage outside of 2 0 containment 0 3 1 1

5. Emergency 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the Director Judgment Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Diredtor that indicates Potential Loss el te Containment Loss of the Fuel Clad Bater. Potential Loss of the Fuel Clad Barrier. Loss of the RCS Baoer. Potential Loss of the RCS Bater. Loss of the Containment Baoier. Boatrer.

Page 12 of 66

NEI 99-01 Rev 6 Proposed EAL Justification F~l[F, jJ- No Change [j Difference Li Deviation Category: Fuel Clad Barrier Category: Fuel Clad Barrier RCS or SG tube leakage RCS or SG tube leakage 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, 3,4 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss Potential Loss A. RCS/reactor vessel level less than (site-specific level). 1. RCITS hot leg instruments indicate 0 inches after lowering trend.

AND

2. In-core thermocouples are unavailable.

AND

3. ALL RCP's are secured.

Page 13 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC2 Category: Fuel Clad BarrierC2 No Change D Difference Deviation Inadequate Heat Removal RCS Activity 1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability: 2) Potential Loss #3, The initiation of HPI - PORV cooling creates a controlled Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3, 4 opening of the RCS to the RB by an open PORV. This is indicative of the steam Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: generators inability to remove heat from the RCS and represents a potential challenge to the FC barrier and is considered a potential loss.

Loss Loss A. Core exit thermocouple readings greater than (site-specific temperature value). 1. Trad > 1400°F Potential Loss Potential Loss A. Core exit thermocouple readings greater than (site-specific temperature value). 2. > 25°F Superheat OR OR B. Inadequate RCS heat removal capability via steam generators as indicated by 3. HPI-PORV Cooling in effect.

(site-specific indications).

Page 14 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC3 Category: Fuel Clad Barrier 'C3 E No Change Difference Deviation Containment Radiation / RCS Activity Containment Radiation / RCS Activity 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. 4 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Containment radiation monitor reading greater than (site-specific value). 1. Containment radiation monitor (RM-G-22 or RM-G-23) reading OR > 1.95E+03 R/hr.

OR B. (Site-specific indications that reactor coolant activity is greater than 300 iC[igm 2. Coolant activityR> 300uCigm Dose Equivalent 1-131 dose equivalent 1-131).

Page 15 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Fuel Clad Barrier FC6 Category: Fuel Clad Barrier FC5 No Change Differne Deviation Emergency Director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup. Hot Standby, Hot Shutdown 1,2, 3, 4 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 RC1 C1 [--]No Change [---- Difference i* Deviation Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier RCS or SG Tube Leakage RCS or SG Tube Leakage 1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability: 2) Potential loss threshold #2 is based on the inability to maintain normal liquid Power Operation, Startup, Hot Standby, Hot Shutdown 1,2, 3.4 inventory within the RCS by normal operation of the Make Up System, when one Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: Make Up Pump is discharging to the charging header. The need for a second Make Up Pump or the use of the high capacity makeup rate would be indicative of Loss Loss a substantial RCS leak. 120 gpm is the nominal capacity of each Make Up Pump.

However, 150 gpm (high makeup flow alarm setpoint) was selected because it is A. An automatic or manual ECCS (SI) actuation is required by EITHER of the 1. Automatic or manual ESAS actuation is required by EITHER of the following: more easily recognized by the operator which will result in a more timely following: a. UNISOLABLE RCS leakage declaration than performing a calculation.

1. UNISOLABLE RCS leakage OR OR b. Steam Generator tube RUPTURE. 2) Potential Loss #3, The initiation of HPI - PORV cooling creates a controlled

2. SG tube RUPTURE. Potential Loss opening of the RCS to the RB by an open PORV. The opening of the RCS and the sPotential Loss pressure control mode through throttling of the HPI and or MU valves represents a

2. UNISOLABLE RCS leakage > 150gpm. potential challenge to the RCS barrier and is considered a potential loss.

A. Operation of a standby charging (makeup) pump is required by EITHER of the OR following: 3) Potential Loss #4, The pressurizer code safety valves will open between 2450

1. UNISOLABLE RCS leakage 3. HPI-PORV Cooling in effect. psig and 2510 psig(allowance for set pressure and Code Safety valve OR OR accumulation). This is the design limit for the RCS and well within tested values (2750 psig). If the RCS heatup is able to keep the pressurizer code safety valves

2. SG tube leakage. 4. a. RCS Pressure > 2450 psig. open with pressure either increasing or cycling then the RCS shall be considered a OR AND potential breach.

B. RCS cooldown rate greater than (site-specific pressurized thermal shock criteria/limits defined by site-specific indications). b. RCS Pressure not loweuing. 4) The proposed EAL provides the Operators with a clear and easily recognizable entry condition for this EAL without altering the intent of the EAL. The usage of the specified thresholds would ensure timely declaration should this event occur.

TMI is using these conditions as entry into this EAL to mimic the concerns of the Potential Loss of RCS as shown in the Westinghouse CSFT monitoring for pressurized thermal shock (PTS) through the use of the RCS integrity red path.

These are the same thresholds as what was approved in EArs based on NEI 99-01 Rev 5.

Page 17 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Category: Reactor Coolant System Barrier RC2 Category: Reactor Coolant System BarrierRC2 No Change Difference Deviation Inadequate Heat Removal Inadequate Heat Removal 1) Listed site-specific threshold value to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability: 2) The initiation of HPI - PORV cooling creates a controlled opening of the RCS to Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 the RB by an open PORVWThe opening of the RCS and the pressure control mode Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: through throttling of the HPI and or MU valves represents a potential challenge to the RCS barrier and is considered a potential loss.

Potential Loss Potential Loss A. Inadequate RCS heat removal capability via steam generators as indicated by HPI-PORV Cooling in effect.

(site-specific indications).

Page 18 of 66

NEI 99-01 Rev 6 Proposed EAL Justification 4 i-RC3 RC3 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change I Difference 1 Deviation Containment Radiation I RCS Activity 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 1,2,3,4 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. Containment radiation monitor reading greater than (site-specific value). Containment radiation monitor (RM-G-22 or RM-G-23) reading > 25 RJhr.

Page 19 of 66

NEt 99-01 Rev 6 Proposed EAL Justification RC6 Category: Reactor Coolant System Barrier RC5 No Change Difference Deviation Category: Reactor Coolant System Barrier Emergency Director Judgment Emergency director Judgment Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,23,34 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 20 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT1 Category: Containment Barrier No Change Diffe ce Deviation Category: Containment Barrier RCS or SG Tube Leakage RCS or SG Tube Leakage 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, 4 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Loss Loss A. A leaking or RUPTURED SG is FAULTED outside of containment. 1. SG tube leakage > 150gpm AND

2. SG is FAULTED outside of containment.

Page 21 of66

Justification Proposed EAL NEI NEI 99-01 Rev 66 99-01 Rev Proposed EAL Justification i +i CT2 CT2 Category: Containment Barrier Category: Containment Barrier I--E No Change D: Difference 1:1 Deviation Inadequate Heat Removal Inadequate Heat Removal 1) Listed site-specific threshold values to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss Potential Loss A. 1. (Site-specific criteria for entry into core cooling restoration procedure)

1. T.,ý > 1 800°F.

AND AND

2. Restoration procedure not effective within 15 minutes. 2. EOP Restoration procedures not effective in < 15 minutes.

Page 22 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT3 Category: Containment Barrier CT3 [ No Change Differce Deviation Category: Containment Barrier Containment Radiation / RCS Activity Containment Radiation / RCS Activity 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, 4 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:

Potential Loss Potential Loss A. Containment radiation monitor reading greater than (site-specific value). Containment radiation monitor (RM-G-22 or RM-G-23) reading > 4.40E+03 R/hr.

Page 23 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT4 CT4 - NChne Dfeee Deato Category: Containment Barrier C No Change Difference Deviation Category: Containment Barrier Containment Integrity or Bypass Containment Integrity or Bypass 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, 4 Fission Product Barrier Threshold: Fission Product Barrier(FPB) Threshold:

Loss Loss A. Containment isolation is required 1. Containment isolation is required and EITHER of the following:

AND a. UNPLANNED lowering in containment pressure or rise in radiation monitor readings outside of containment in the Emergency Directors judgment EITHER of the following: indicate a loss of containment integrity.

1. Containment integrity has been lost based on Emergency Director OR judgment.

OR b. UNISOLABLE pathway from containment to the environment exists.

OR

2. UNISOLABLE pathway from the containment to the environment exists.

OR 2. Indication of RCS leakage outside of containment B. Indications of RCS leakage outside of containment. Potential Loss Potential Loss 3. Reactor Building Pressure > 55 psig and rising.

OR A. Containment pressure greater than (site-specific value) 4. Hydrogen Concentration in Containment > 4%.

OR OR B. Explosive mixture exists inside containment 5. a. Reactor Building pressure > 30 psig OR AND C. 1. Containment pressure greater than (site-specific pressure setpoint) b. Reactor Building Emergency cooling is less than ANY one of the following AND conditions:

2. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer. SPRAY COOLERS 2 0 0 3 1 1 Page 24 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CT6 Category: Containment BarriCerT5 No Change Difference 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, 4 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 Barnier.

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 Containment Barrier. Loss of the Containment Barrier.

Page 25 of 66

0 NEI 99-01 Rev 6 1 Proposed EAL I Justification SG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition:

MG1 F- No Change Difference Deviation 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: coping analysis, and site specific indication to ensure timely classification.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels:

Emergency Action Levels (EAL):

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

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 Emergency 4KV buses.

AND AND 2. Failure of EG-Y-1A, EG-Y-1B Emergency Diesel Generators and EG-Y-4 SBO Diesel Generator to supply power to Emergency 4KV buses.

b. EITHER of the following:

AND

" Restoration of at least one emergency bus in less than 3. EITHER of the following:

(site-specific hours) is not likely. a. Restoration of at least one Emergency 4KV bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely.

OR

" (Site-specific indication of an inability to adequately remove heat from the core) b. > 25°F superheat Page 26 of 66

0 NEt 99-01 Rev 6 fProposed EAL Justification ss1 MS1 Initiating Condition: SITE AREA EMERGENCY Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Initiating Condition: H-* No Change 1-1 Difference FIDeviation Loss of all offsite and onsite AC power to emergency busses for 15 minutes or 1) Listed site specific equipment to ensure timely classification.

longer.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels:

Emergency Action Levels (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 determining that the applicable time has been exceeded, or will likely be exceeded.

Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 minutes or longer. 1. Loss of ALL ofasite AC Power to Emergency 4KV buses.

AND

2. Failure of EG-Y-1A, EG-Y-1B Emergency Diesel Generators and EG-Y-4 SBO Diesel Generator to supply power to Emergency 4KV buses.

AND

3. Failure to restore power to at least one Emergency 4KV bus in < 15 minutes from the time of loss of both offsite and onsite AC power Page 27 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SA1 MAI Initiating Condition: ALERT Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Initiating Condition: M No Change FIDifference I] Deviation 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:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels:

Emergency Action Levels (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 determining that the applicable time has been exceeded, or will likely be

1. a. AC power capability to (site-specific emergency buses) is reduced to a. exceeded.

single power source for 15 minutes or longer.

1. AC power capability to Emergency 4KV buses reduced to only one of the AND following power sources for > 15 minutes.

b. Any additional single power source failure will result in loss of all AC

Auxiliary Transformer 1A power to SAFETY SYSTEMS. Auxiliary Transformer 1B

Emergency Diesel Generator EG-Y-1A

Emergency Diesel Generator EG-Y-1B

SBO Diesel Generator EG-Y-4 AND

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

.1 I Page 28 of 66

NEI 99-01 Rev 6 1 Proposed EAL I Justification Sul MUl Initiating Condition: UNUSUAL EVENT Initiating Condition: I No Change FIDifference I Deviation Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels:

Emergency Action Levels (EAL):

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

Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 minutes or longer Loss of ALL offsite AC power capability to Emergency 4KV busses for a 15 minutes.

Page 29 of 66

NEI 99-01 Rev 6 1 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 -IDeviation Loss of all AC and Vital DC power sources 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 Power Operation. Startup, Hot Standby, Hot Shutdown cause confusion on the need to declare, 1,2,3,4 Example Emergency Action Levels:

Emergency Action Levels (EAL):

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

1. 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 Emergency 4KV buses.

AND AND

2. Failure of EG-Y-1A, EG-Y-1B Emergency Diesel Generators and EG-Y-4 SBO Indicated voltage is less than (site-specific bus voltage value) on ALL Diesel Generator to supply power to Emergency 4KV buses.

(site-specific vital DC busses) for 15 minutes or longer.

AND

3. Voltage is < 105 VDC on 125 VDC Distribution System 1A and lB.

AND

4. ALL AC and Vital DC power sources have been lost for >_15 minutes.

Page 30 of 66

NEI 99-01 Rev 6 JProposed EAL Justification b$8 MS2 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: D No Change I Difference ] 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 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 Power Operation, Startup, Hot Standby, Hot Shutdown cause confusion on the need to declare.

1,2,3,4 Example Emergency Action Levels:

Emergency Action Levels (EAL):

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

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

Voltage is < 105 VDC on 125 VDC Distribution System 1A and lB for

>15 minutes.

Page 31 of`66

NEI 99-01 Rev 6 1 Proposed EAL I Justification SS5 Initiating Condition: SITE AREA EMERGENCY Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPV Initiating Condition:

MS3 D- No Change F Difference [IDeviation water level [BWR]) or RCS heat removal. Inability to shutdown the reactor causing a challenge to core cooling 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:

3) The initiation of HPI - PORV cooling creates a controlled opening of the RCS to Power Operation the RB by an open PORV. This is indicative of the steam generators inability to 1.2 remove heat from the RCS.

Example Emergency Action Levels:

Emergency Action Levels (EAL):

1. a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the reactor. 1. Automatic or Manual Trip 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 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:

a. Tad > 1400°F.

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

b. HPI-PORV Cooling in effect.

OR

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

Page 32 of 66

NEI 99-01 Rev 6 Justification

_A Prpoe NEt 99-01 Rev B i Proposed EAL Justification SA5 MA3 Initiating Condition: ALERT Automatic or manual (trip [PWRJ / scram [BWRJ) fails to shutdown the reactor, and Initiating Condition: F-1 No Change FIDifference 1 Deviation subsequent manual actions taken at the reactor control consoles are not Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down 1) Listed site specific indications to ensure timely classification.

successful in thutting down the reactor.

the reactor.

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 Levels (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: A manual action is any operator action, or set of actions, which causes the driving in control rods or implementation of boron injection strategies. 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. An automatic (trip [PWR] / scram [BWR]) did not shutdown the

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

> 5%.

AND AND

2. Manual actions taken at the Console Center are not successful in shutting

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

successful in shutting down the reactor.

Page 33 of 66

NEI 90-01 Rev 6 [Proposed EAL Justification SUS MU3 Initiating Condition: UNUSUAL EVENT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor.

Initiating Condition: F- No Change I-1 Difference FIDeviation Automatic or manual trip 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 Levels (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: A manual action is any operator action, or set of actions, which causes the driving in control rods or implementation of boron injection strategies. 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. An automatic (trip [PWR] / scram [BWRI) did not shutdown the reactor. 1. a. Automatic Trip did not shutdown the reactor as indicated by Reactor Power > 5%.

AND AND

b. Subsequent manual action taken at the Console Center is successful in

b. A subsequent manual action taken at the reactor control consoles is shutting down the reactor.

successful in shutting down the reactor.

OR

2. a. Manual Trip did not shutdown the reactor as indicated by Reactor Power

2. a. A manual trip ([PWR] / scram [BWR]) did not shutdown the reactor.

> 5%.

AND AND

b. Subsequent automatic Trip is successful in shutting down the reactor.

b. EITHER of the following:

1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.

OR

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

Page 34 of 66

NEI 99-01 Rev 6 Proposed EAL Justification

-F +

SA2 Initiating Condition: ALERT UNPLANNED lost, of Control Room indications for 15 minutes or longer with a significant Initiating Condition:

MA4 H No Change FIDifference -IDeviation transient in progress. UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. 1) Listed site specific number of steam generators to ensure timely classification.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1.2,23,4 Example Emergency Action Levels:

Note: The Emergency Director should declare the Alert promptly upon determining that Emergency Action Levels (EAL):

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 following determining that the applicable time has been exceeded, or will likely be parameters from within the Control Room for 15 minutes or longer. exceeded.

[see table below]

1. a An UNPLANNED event results in the inability to monitor ANY Table M1 I BWR parameter list] I PWR parameter list] parameters from within the Control Room for > 16 minutes.

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

Reactor Power

PZR Level RPV Pressure RCS Pressure

RCS Pressure

In Core/Core Exit Temperature Primary Containment Pressure In Core/Core Exit Temperature

Level in at least one OTSG.

Suppression Pool Level Levels in at least (site specific number) steam generators

OTSG Emergency Feed Water Flow Steam Generator Auxiliary or Suppression Pool Temperature Emergency Feed Water Flow AND AND b. ANY Table M2 transient in progress.

b. Any of the following transient events in progress.

Table M2 Significant Transients

Automatic or Manual runback greater than 25% thermal reactor power

Electrical load rejection greater than 25% full electrical load

Automatic Turbine Runback >25% thermal reactor power Reactor Scram [BWR] / trp [PWR]

ECCS (SI) actuation

Electrical Load Rejection >26% full electrical load

Thermal power oscillations greater than 10% [BWR]

Reactor Trip

ESAS Actuation

Thermal Power oscillations > 10%

Page 35 of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification SU2 MU4 Initiating Condition: UNUSUAL EVENT Initiating Condition: F No Change 1 Difference FI Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer.

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

1) Listed site specific number of steam generators to ensure timely classification.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels:

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

determining that 15 minutes has been exceeded, or will likely be exceeded. Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be An UNPLANNED event results in the inability to monitor one or more of the exceeded.

following parameters from within the Control Room for 15 minutes or longer.

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

[ BWR parameter list] [PWR parameter list]

Table M1 Control Room Parameters Reactor Power Reactor Power

" Reactor Power RPV Level RCS Level " PZR Level

RCS Pressure RPV Pressure RCS Pressure

In Core/Core Exit Temperature Prmary Containment Pressure In Core/Core Exit Temperature " Level in at least one OTSG.

nPool Level Levels in at least (site specific " OTSG Emergency Feed Water Flow Suppression Pnumber) steam generators Steam Generator Auxiliary or Suppression Pool Temperature Emergency Feed Water Flow Page 36 of 66

0 NEI 99-01 Rev 6 Proposed EAL j Justification SA9 MA5 Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM needed for the current Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM required for the current F] No Change M Difference [Deviation operating mode. operating mode. 1) No additional site specific hazards noted 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 operators and Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 minimize confusion.

Example Emergency Action Levels: Emergency Action Levels (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

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

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

OR OR

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

Page 37 of 66

NEI 99-01 Rev 6 Proposed EAL Justification SU4 MU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change 1 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:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels: (1 or 2 or 3)

Emergency Action Levels (EAL):

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

1. RCS unidentified or pressure boundary leakage greater than (site-specific value) for 15 minutes or longer. 1. RCS unidentified or pressure boundary leakage > 10 gpm for > 15 minutes

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

2. RCS identified leakage >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 containment >28 gpm for

> 15 minutes Page 38 of 66

NEI 99-01 Rev 6 ]Proposed EAL I Justification SU6 MU7 InitiatingCondition:UNUSUAL EVENT Initiating Condition: F No Change FIDifference -IDeviation Loss of all onsite or offsite communications capabilities Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

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

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3,4 Example Emergency Action Levels: (1 or 2 or 3)

Emergency Action Levels (EAL):

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

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

2. Loss of ALL of the following ORO communications s) methods: OR (site-specific list of communications methods) 2. Loss of ALL Table M3 Offsite communication capability affecting the 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 M3 NRC communication capability affecting the ability to perform NRC notifications.

Table M3 Communications Capability System Onsite Offsite NRC Radios X Plant page X Plant Telephone System X Sound Powered Phones X Commercial Telephones X X X NARS X ENS X X HPN X X Satellite phones X X Page 39 of 66

NEI 99-01 Rev 6 1 Proposed EAL [Justification SU7 MU8 Initiating Condition: UNUSUAL EVENT Initiating Condition: FH No Change 1 -]Difference FIDeviation Failure to isolate containment or loss of containment pressure control. [PWR]

Failure to isolate containment or loss of containment pressure control.

1) Listed site specific indications to ensure timely classification.

Operating Mode Applicability:

Operating Mode Applicability: 2) Reworded EAL 1.b to be a positive statement Power Operation, Startup, Hot Standby, Hot Shutdown 1,2.3,4 Example Emergency Action Levels: (1 or 2)

Emergency Action Levels (EAL):

1. a. Failure of containment to isolate when required by an actuation signal. 1. a. Failure of containment to isolate when required by an actuation signal.

AND AND

b. ANY required penetration remains open > 15 minutes of the actuation

b. ALL required penetrations are not closed within 15 minutes of the signal.

actuation signal. OR

2. a. Reactor Building pressure > 30 psig

2. a. Containment pressure greater than (site-specific pressure).

AND AND b. Reactor Building Emergency cooling is less than ANY one of the following conditions for > 15 minutes:

b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.

SPRAY COOLERS 2 0 0 3 1 1 Page 40 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CA2 CA1 Initiating Condition: ALERT Initiating Condition:

E No Change 1 Difference 1 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 busses for 15 minutes or 1) Listed site specific equipment to ensure timely classification.

longer. longer.

Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown, Refueling, Defueled 5,6, D Example Emergency Action Levels: Emergency Action Levels (EAL):

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

1. Loss of ALL offsite AC power to Emergency 4KV buses.

Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses)

AND for 15 minutes or longer.

2. Failure of EG-Y-1A, EG-Y-1B Emergency Diesel Generators and EG-Y-4 SBO Diesel Generator to supply power to Emergency 4KV buses.

AND

3. Failure to restore power to at least one Emergency 4KV bus in < 15 minutes from the time of loss of both offsite and onsite AC power.

Page 41 of 66

Justification Proposed EAL NEI Rev' 6S NEt 99-01 Rev Proposed EAL Justification

+ -t CU2 Cul E No Change 1 Difference 1 Deviation Initiating Condition: UNUSUAL EVENT Initiating Condition:

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:

Cold Shutdown, Refueling. Defueled 5.6,0 Example Emergency Action Levels:

Emergency Action Levels (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 determining that the applicable time has been exceeded, or will likely be

1. a. AC power capability to (site-specific emergency buses) is reduced to a exceeded.

single power source for 15 minutes or longer.

AND 1. AC power capability to Emergency 4KV buses reduced to only one of the

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

power to SAFETY SYSTEMS.

" Auxiliary Transformer 1A

" Auxiliary Transformer 1 B

" Emergency Diesel Generator EG-Y-1A

" Emergency Diesel Generator EG-Y-1B

" SBO Diesel Generator EG-Y-4 AND

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

Page 42 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition - ALERT CA6 Initiating Condition:

CA2 H No Change E Difference Deviation Hazardous event affecting SAFETY SYSTEM needed for the current operating Hazardous event affecting SAFETY SYSTEM required for the current operating mode. mode. 1) No additional site specific hazards noted Operating Mode Applicability: Operating Mode Applicability:

2) Changed the word "needed" to "required" in the IC and "required by Cold Shutdown, Refueling 5,6 Technical Specification" in the EAL to be consistent with terminology used by operators and minimize confusion.

Example Emergency Action Levels: Emergency Action Levels (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 tomado 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 required by Technical SYSTEM component or structure needed for the current Specifications for the current operating mode.

operating mode.

Page 43 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CU4 CU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: D- No Change [-q Difference [I] Deviation Loss of Vital DC power for 15 minutes or longer. Loss of Vital DC power for 15 minutes or longer.

Operating Mode Applicability: Operating Mode Applicability: 1) Listed site specific voltage and equipment to ensure timely classification.

Cold Shutdown, Refueling 5, 6 2) Removed the word "indicated" this will allow for an indication problem to not cause confusion on the need to declare.

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

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

Indicated voltage is less than (site-specific bus voltage value) on required Vital DC Voltage is < 105 VOC on required 125 VDC Distribution System 1A and 1B for buses for 15 minutes or longer. a 15 minutes.

Page 44 of 66

NEI 99-01 Rev 6 Proposed EAL Justification CU5 CU4 Initiating Condition: UNUSUAL EVENT Initiating Condition: E No Change 1 Difference 1 Deviation Loss of all ornsite or offsite communications capabilities Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

1) Listed site specific communications methods to ensure timely classification Operating Mode Applicability:

Cold Shutdown, Refueling, Defuled 5,6, D Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Levels (EAL):

1. Loss of ALL of the following onsite communication methods: 1. 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 C1 Communications Ca pability System Onsite Offsite NRC Radios X Plant page X Plant Telephone System X Sound Powered Phones X Commercial Telephones X X X NARS X ENS X X HPN X X Satellite phones X X Page 45 of 66

NEI 99-01 Rev 6 Proposed EAL ]Justification CA3 CA5 Initiating Condition: ALERT Initiating Condition: 1:1 No Change El Difference [--] 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 to Operating Mode Applicability: Operating Mode Applicability: ensure timely classification.

Cold Shutdown, Refueling 5, 6

2) Listed site specific pressure reading to enhance timely classification.

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

Note: The Emergency Director should declare the Alert promptly upon Note: The Emergency Director should declare the event promptly upon 3) Added wording relating the temp and press rise to a loss of decay heat removal determining that the applicable has been exceeded, or will likely be determining that the applicable time has been exceeded, or will likely be capability as per the developer notes for PWR's exceeded. exceeded.

0

1. UNPLANNED increase in RCS temperature to greater than (site-specific 1. UNPLANNED rise in RCS temperature > 200 F due to loss of decay Technical Specification cold shutdown temperature limit) for greater than heat removal for > Table C2 duration.

the duration specified in the following table.

OR

2. UNPLANNED RCS pressure increase greater than (site-specific pressure reading). (This EAL does not apply during water-solid plant conditions.

2. UNPLANNED RCS pressure rise > 10 psig as a result of temperature

[PWR])

rise due to loss of decay heat removal. (This EAL does not apply during water- solid plant conditions.)

Table: RCS Heat-up Duration Thresholds Thht~ C2 RC~ Weat-,~mO.,rattnn Threchnids RCS Status Containment Closure Heat-up Duration RCS Status Containment Closure Heat-up Duration Status Status Intact (but not Intact Not Applicable 60 minutes*

RCS Reduced Not Applicable 60 minutes*

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

reduced inventory - OR

[PWRJ) Not Established 0 minutes

If an RCS heat removal system is in operation within this time frame and Reduced RCS temperature is being reduced, the EAL is not applicable. 0 minutes Inventory Not Established If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, then EAL #1 is not apolicable.

Page 46 of 66

NPI 99-01 Rev 6 Proposed EAL Justification CU3 CU5 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change E Difference FIDeviation UNPLANNED increase in RCS temperature. UNPLANNED rise in RCS temperature. 1) Listed site specific Technical Specification cold shutdown temperature limit Operating Mode Applicability: to ensure timely classification.

Operating Mode Applicability:

Cold Shutdown, Refueling 5,6 2) Changed the word increase to rise in the initiating condition to be consistent Example Emergency Action Levels: with operations language and training.

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

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

1. UNPLANNED increase in RCS temperature to greater than (site-specific 1. UNPLANNED rise in RCS temperature > 200*F due to loss of decay Technical Specification cold shutdown temperature limit). heat removal.

2. Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV OR

[BWRI) level indication for 15 minutes or longer.

2. Loss of the following for > 15 minutes.

ALL RCS temperature indications AND

ALL RCS level indications Page 47 of 66

NEI 99-01 Rev 6 1 Proposed EAL Justification CG1 CG6 Initiating Condition: GENERAL EMERGENCY Initiating Condition: H No Change E Difference 1 Deviation Loss of (reactor vessel/RCS IPWR] or RPV [BWRI) inventory affecting fuel clad Loss of reactor vessel / RCS inventory affecting fuel clad integrity with integrity with containment challenged. containment challenged. 1) EAL 1a not included as per guidance in developer notes since top of active fuel is below level indication lowest value.

Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown. Refueling 5,6 2) Listed site specific radiation monitors to ensure timely classification Example Emergency Action Levels: (1 or 2) Emergency Action Levels (EAL):

3) Listed site specific sumps and tanks to ensure timely classification Note: The Emergency Director should declare the Unusual Event promptly Note: The Emergency Director should declare the event promptly upon upon determining that 30 minutes time has been exceeded, or will likely determining that the applicable time has been exceeded, or will likely be 4) Listed Explosive mixture in the Containment Challenge Table to ensure be exceeded. exceeded. timely classification

1. a. (Reactor vessel/RCS [PWRJ or RPV [BWRJ) vessel level less than (site- 1. Reactor Vessel / RCS level unknown for >30 minutes. 5) Worded "cannot be monitored" as unknown to ensure clarity for instances specific level) for 30 minutes or longer. when the indicator is working but is over/under ranged. This is also in keeping AND AND with current EAL wording.

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

b. ANY indication from the Containment Challenge Table

Table C3 indications of a sufficient magnitude to indicate core

2. a.. (Reactor vessel/RCS [PWR] or RPV [BWR]) vessel level cannot be uncovery monitored for 30 minutes or longer. OR AND

Erratic Source Range Neutron Monitor indication.

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

OR

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

Radiation Monitor RM-G-6 or RM-G-7 reading > 3 R/hr.

Erratic source range monitor indication [PWRJ AND

UNPLANNED increase in (site-specific sump and/or tank levels) of 3. ANY Containment Challenge Indication (Table C4) sufficient magnitude to indicate core uncovery Table C3 Indications of RCS Leakage

(Other site-specific indications)

AND

UNPLANNED Reactor Building Sump level rise*

UNPLANNED Auxiliary Bldg Sump level rise'

c. ANY indication from the Containment Challenge Table).

UNPLANNED BWST level rise'

UNPLANNED RCDT level rise*

Table: Containment Challenge Table

UNPLANNED rise in RCS makeup

" CONTAINMENT CLOSURE not established*

Observation of leakage or inventory loss

(Explosive mixture) exists inside containment .Rise in level is attributed to a loss of reactor vessel/RCS inventory.

UNPLANNED increase in containment pressure

Secondary containment radiation monitor reading above (site-specific value) [BWRI Table C4 Containment Challenge Indications

'if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is " Hydrogen Concentration in Containment > 4%

not required.

UNPLANNED rise in containment pressure

" CONTAINMENT CLOSURE not established'

'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

NEI 99-01 Rev 6 1 Proposed EAL ] Justification Initiating Condition: SITE AREA EMERGENCY CS1 Initiating Condition:

CS6

[ No Change E Difference 1 Deviation Loss of (reactor vesselIRCS [PWR] or RPV [BWR]) inventory affecting core Loss of reactor vessel / RCS inventory affecting core decay heat removal 1) EAL 1 not included as per guidance in developer notes since 6" below decay heat removal capability. capabilities. bottom ID of RCS loop is below level indication lowest value.

Operating Mode Applicability: Operating Mode Applicability:

2) EAL 2 not included as per guidance in developer notes since top of active Cold Shutdown, Refueling 5,6 fuel is below level indication lowest value,.

Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Levels (EAL): 3) Listed site specific radiation monitors to ensure timely classification Note: The Emergency Director should declare the Unusual Event promptly Note: The Emergency Director should declare the event promptly upon upon determining that 30 minutes time has been exceeded, or will likely determining that the applicable time has been exceeded, or will likely 4) Listed site specific sumps and tanks to ensure timely classification be exceeded. be exceeded.

5) Worded "cannot be monitored" as unknown to ensure clarity for instances

1. a. CONTAINMENT CLOSURE not established. when the indicator is working but is over/under ranged. This is also in keeping AND 1. Reactor vessel level unknown for >30 minutes. with current EAL wording.

b. (Reactor vessel/RCS [PWR) or RPV [BWR]) level less than (site- AND specific level).

2. Core uncovery is indicated by any of the following:

2. a. CONTAINMENT CLOSURE established.

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

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

Erratic Source Range Neutron Monitor indication.

3. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be OR monitored for 30 minutes or longer.

AND

Radiation Monitors RM-G-6 or RM-G-7 reading > 3 R/hr.

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

Table C3 Indications of RCS Leakage

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

UNPLANNED Reactor Building Sump level rise*

Erratic source range monitor indication [PWR]

UNPLANNED Ausliary Bldg. Sump level rise'

UNPLANNED BWST level rise'

UNPLANNED increase in (site-specific sump and/or tank

UNPLANNED RCDT level rise*

levels) of sufficient magnitude to indicate core uncovery

UNPLANNED rise in RCS makeup

(Other site-specific indications)

Observation of leakage or inventory loss

'Rise in level is attributed to a loss of reactor vesselIRCS inventory.

Page 49 of 66

NEI 99-01 Rev 6 Proposed EAL Justification t*AI CAB Initiating Condition: ALERT Initiating Condition: D No Change I Difference L-- Deviation Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory Loss of reactor vessel / RCS inventory

1) Listed site specific levels to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Cold Shutdown, Refueling 5, 6 2) Listed site specific sumps and tanks to ensure timely classification.

Example Emergency Action Levels: (1 or 2) Emergency Action Levels (EAL): 3) Worded "cannot be monitored" as unknown to ensure clarity for instances Note: The Emergency Director should declare the event promptly upon when the indicator is working but is over/under ranged. This is also in keeping Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes time has been exceeded, or will likely determining that the applicable time has been exceeded, or will likely with current EAL wording.

be exceeded. be exceeded.

1. Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as 1. Loss of Reactor Vessel / RCS inventory as indicated by RCS level indicated by level less than (site-specific level). < 0 inches on Draindown Level indicator.

2. a. (Reactor vessel/RCS [PWRI or RPV [BWR]) level cannot be OR monitored for 15 minutes or longer 2. a. Reactor vessel / RCS level unknown for > 15 minutes.

AND AND

b. UNPLANNED increase in (site-specific sump and/or tank) levels due b. Loss of reactor vessel / RCS inventory per Table C3 indications.

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

Table C3 Indications of RCS Leakage

UNPLANNED Reactor Building Sump level rise*

UNPLANNED AuJliary Bldg. Sump level rise*

UNPLANNED BWST level rise*

UNPLANNED RCDT level rise'

UNPLANNED rise in RCS makeup

Observation of leakage or inventory loss Rise in level is attributed to a loss of reactor vessel/RCS inventory.

Page 50 of 66

NEI 99-01 Rev 6 Proposed EAL Justification Cu1 CU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: El No Change I Difference FIDeviation UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory for UNPLANNED loss of reactor vessel / RCS inventory for 15 minutes or longer.

15 minutes or longer. 1) Described "a required lower limit" as a procedurally established lower limit, Operating Mode Applicability:

Operating Mode Applicability:

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

5, 6 Cold Shutdown, Refueling 2) Worded "cannot be monitored" as unknown to ensure clarity for instances Emergency Action Levels (EAL):

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

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

exceeded.

1. UNPLANNED loss of reactor coolant results in the inability to restore

1. UNPLANNED loss of reactor coolant results in (reactor vessel/RCS and maintain reactor vessel / RCS level to

[PWR] or RPV [BWR]) level less than a required lower limit for 15 > procedurally established lower limit for >_15minutes.

minutes or longer.

OR

2. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be

2. a. Reactor vessel / RCS level unknown.

monitored.

AND AND

b. Loss of reactor vessel / RCS inventory per Table C3 indications.

b. UNPLANNED increase in (site-specific sump and/or tank) levels.

Table C3 Indications of RCS Leakage

UNPLANNED Reactor Building Sump level rise*

UNPLANNED Auxiliary Bldg. Sump level rise*

UNPLANNED BWST level rise"

UNPLANNED RCDT level rise*

UNPLANNED rise in RCS makeup

Observation of leakaoe or inventory loss

Rise in level is attrbuted to a loss of reactor vessel/RCS inventory.

Page 51 of 66

NEI 99-01 Rev 6 Proposed EAL Justification HG1 HG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition: D-- No Change F Difference Deviation

1) List site security shift supervision as Security Force.

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

2) Added descriptors to better explain each safety function and allow for a Operating Mode Applicability: Operating Mode Applicability: timely classification.

AJl 1, 2, 3,4, 5, 6, D Example Emergency Action Levels: Emergency Action Levels (EAL):

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

supervision).

AND AND

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

b. EITHER of the following:

OR

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

b. Damage to spent fuel has occurred or is IMMINENT Reactivity control Core cooling [PWR] I RPV water level [BWRJ

RCS heat removal Table H1 Safety Functions OR " Reactivity Control (ability to shut down the reactor and keep it shutdown)

2. Damage to spent fuel has occurred or is IMMINENT " Core Cooling (ability to cool the core)

" RCS Heat Removal (ability to maintain heat sink)

Page 52 of 66

0 NEI 99-01 Rev 6 Proposed EAL I Justification HS1 HSI Initiating Condition: SITE AREA EMERGENCY Initiating Condition: E No Change FIDifference [- 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, 6, D Example Emergency Action Levels: Emergency Action Levels (EAL):

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

Page 53 of 66

NEI 99-01 Rev 6 JProposed EAL Justification HA1 HA1 Initiating Condition: ALERT Initiating Condition: F-x No Change FIDifference F Deviation HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack HOSTILE ACTION within the OWNER CONTROLLED AREA or airbome attack threat within 30 minutes. threat within 30 minutes. 1)Lisl site security shift supervision as Security Force.

Operating Mode Applicability: Operating Mode Applicability:

All 1,2, 3,4,5, 6, D Example Emergency Action Levels: (1 or 2) Emergency Action Levels (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).

OR

2. A validated notification from NRC of an aircraft attack threat within 30 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 HUI Initiating Condition: UNUSUAL EVENT Initiating Condition:

HU1 F No Change -IDifference LI1 Deviation Confirmed SECURITY CONDITION or threat.

Confirmed SECURITY CONDITION or threat.

1) List site security shift supervision as Security Force.

Operating Mode Applicability:

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

AJI 1,2, 3,4, 5, 6, D Example Emergency Action Levels: (1 or 2 or 3)

Emergency Action Levels (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.

OR

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

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

3. A validated notification from the NRC providing information of an aircraft 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 Initiating Condition: SITE AREA EMERGENCY HS6 Initiating Condition:

HS2 FD No Change M Difference 1 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 procedures to effectively list all of the alternate locations, panels, and stations requested by Operating Mode Applicability: Operating Mode Applicability: the developer notes. This would be the procedures the Control Room would 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, 6, D shutdown were to be listed, Example Emergency Action Levels: (1 and 2) Emergency Action Levels (EAL): 2) Added descriptors to better explain each safety function and allow for a Note: The Emergency Director should declure the event promptly upon determining Note: The Emergency Director should declare the event promptly upon determining timely classification.

that (site-specific number of minutes) has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded.

exceeded. 3) 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

1. A Control Room evacuation has resulted in plant control being transferred from the room still manned, due to testing or equipment failure.

1. An event has resulted in plant control being transferred from the Control Control Room to alternate locations per OP-TM-EOP-020, Cooldown from Outside Room to (site-specific remote shutdown panels and local control stations). the Control Room.

AND

2. Control of ANY of the following key safety functions is not reestablished within (site-specific number of minutes). 2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.

Reactivity control

" Core cooling [PWRJ I RPV water level [BWR] Table HI Safety Functions RCS heat removal

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

Core Cooling (ability to cool the core)

RCS Heat Removal (ability to maintain heat sink)

Page 56 of 66

NEt 99-01 Rev 6 1Proposed EAL IJustification HA6 HA2 Initiating Condition: ALERT Initiating Condition: D No Change [F Difference -- ] Deviation Control Room evacuation resulting in transfer of plant control to alternate locations. I Control Room evacuation resulting in transfer of plant control to alternate locations.

1) EAL uses the site specific Control Room evacuation procedures to effectively list all of the alternate locations, panels, and stations requested by the developer Operating Mode Applicability: Operating Mode Applicability: notes. This would be the procedures the Control Room would enter should such an event occur, this allows for greater clarity as to when this EAL would apply than if All 1,2, 3,4, 5, 6, D each panel and station used in alternate shutdown were to be listed,

2) Changed "An event" to" A Control Room evacuation" to remove confusion if Example Emergency Action Levels: Emergency Action Levels (EAL): partial plant control was transferred to outside the control room with the control An event has resulted in plant control being transferred from the Control Room to room still manned, due to testing or equipment failure.

(site-specific remote shutdown panels and local control stations). A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per OP-TM-EOP-020, Cooldown from Outside the Control Room.

Page 57 of 66

NEI 99-01 Rev 6 j Proposed EAL Justification HU4 HU3 Initiating Condition: UNUSUAL EVENT FIRE potentially degrading the level of safely of the plant.

Initiating Condition: LII No Change [-] Difference [] Deviation FIRE potentially degrading the level of safety of the plant.

1) Listed site specific list of plant rooms or areas that contain SAFETY SYSTEM Operating Mode Applicability: equipment to ensure timely classification.

Operating Mode Applicability:

All 1,2, 3,4, 5, 6, D Example Emergency Action Levels: (1 or 2 or 3 or 4)

Emergency Action Levels (EAL):

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

1. a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications: 1. a. A FIRE in any Table H2 area is not extinguished in <15-minutes of ANY

Report from the field (i.e., visual observation) of the following FIRE detection indications:

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

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

Field verification of a single fire alarm

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

Field verification of a single fire alarm

b. The FIRE is located within ANY of the following plant rooms or areas: OR (site-specific list of plant rooms or areas) 2. a. Receipt of a single fire alarm in any Table H2 area (i.e., no other indications of a FIRE).

2. a. Receipt of a single fire alarm (i.e., no other indications of a FIRE). AND AND b. The existence of a FIRE is not verified in <30-minutes of alarm receipt.

b. The FIRE is located within ANY of the following plant rooms or areas: OR (site-specific list of plant rooms or areas) 3. A FIRE within the plant PROTECTED AREA not extinguished in <60-minutes of the initial report, alarm or indication.

AND

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

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

3. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Areal PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. Table H2 Vital Areas

Reactor Building

4. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant

Intake Building Protected Areal PROTECTED AREA that requires firefighting support by an

Intermediate Building offsite fire response agency to extinguish.

Control Tower

Auxiliary and Fuel Handling Buildings

1A and 1B Diesel Generator Buildings

BWST

CST Page 58 of 66

0 NEI 99-01 Rev 6 HU2

[Proposed EAL HU4

]Justification Initiating Condition: UNUSUAL EVENT Initiating Condition: [* No Change 1 Difference FIDeviation Seismic event greater than OBE levels.

Seismic event greater than OBE levels.

1) Listed site specific indication to determining OBE limits have been met or Operating Mode Applicability: exceeded to ensure timely classification.

Operating Mode Applicability:

All 1, 2, 3,4, 5, 6, D Example Emergency Action Levels:

Emergency Action Levels (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) Seismic event > Operating Basis Earthquake (OBE) as indicated by seismic Alarms PRF-1-3 Operating Basis earthquake and PRF-1-2 Threshold Seismic Condition.

Page 59 of 66

NEI W9-01Rev 6 Proposed EAL [Justification HA5 HA5 Initiating Condition: ALERT Initiating Condition: *'* No Change FIDifference FIDeviation Gaseous release impeding access to equipment necessary for normal plant Gaseous release impeding access to equipment necessary for normal plant operations, operations, cooldown or shutdown. cooldown or shutdown. 1) Listed plant specific rooms and areas with entry related mode applicability to ensure timely classification.

Operating Mode Applicability: Operating Mode Applicability:

Ail 1,2, 3,4, 5, 6, D Example Emergency Action Levels: Emergency Action Levels (EAL):

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

1. a. Release of a toxic, corrosive, asphyxiant or flammable gas into any I. Release of a toxic, corrosive, asphyxiant or flammable gas in ANY Table H3 area.

of the following plant rooms or areas:

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

Table H3 Areas with Entry Related Mode Applicability AND Area Entry Related Mode Applicability

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

Reactor Building* Modes 4, 5, and 6 Intermediate Building* Modes 4, 5, and 6 Auxiliary Building* Modes 4, 5, and 6 Fuel Handling Building* Modes 4, 5, and 6

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 HU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: F No Change F Difference FIDeviation Hazardous Event Hazardous Event

1) Included river water level as part of the site-specific list of natural or Operating Mode Applicability: technological hazard events. The EAL values selected are the current Approved Operating Mode Applicability: UE EAL values.

Atl 1, 2, 3,4, 5, 6, D

2) Changed the word "needed" to "required by Technical Specifications" in the EAL Example Emergency Action Levels: (1 or 2 or 3 or 4) to be consistent with terminology used by operators and minimize confusion.

Emergency Action Levels (EAL):

Note: EAL #3 does not apply to routine traffic impediments such as fog, snow.

ice, or vehicle breakdowns or accidents. Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

I. A tornado strike within the PROTECTED AREA.

2. Internal room or area flooding of a magnitude sufficient to require manual 1. Tornado strike within the PROTECTED AREA.

or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. OR

2. Internal room or area flooding of a magnitude sufficient to require manual or

3. Movement of personnel within the PROTECTED AREA is impeded due to automatic electrical isolation of a SAFETY SYSTEM component required by an offsite event involving hazardous materials (e.g., an offsite chemical Technical Specifications for the current operating mode.

spill or toxic gas release).

OR

4. A hazardous event that results in on-site conditions sufficient to prohibit

3. Movement of personnel within the PROTECTED AREA is impeded due to an the plant staff from accessing the site via personal vehicles. offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

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

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 water level at the intake Pump and Screen House, as indicated by EITHER:

a. > 300 ft. el. (high level)

OR

b. < 274 ft. el. (low level)

Page 61 of 66

NEI 99-01 Rev 6 Proposed EAL [Justification HG8 HG7 Initiating Condition: GENERAL EMERGENCY Initiating Condition: x-1 No Change FIDifference FIDeviation 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, 6, D Example Emergency Action Levels: Emergency Action Levels (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 IMMINENT that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action the facility. Releases can be reasonably expected to exceed EPA Protective Guideline exposure levels off-site for more than the immediate site area. Action Guideline exposure levels off-site for more than the immediate site area..

Page 62 of 66

S NEI 99-01 Rev 6 1 Proposed EAL Justification HS8 HS7 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: M No Change. FIDifference 1-- 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:

AJI 1,2,3,4,5,6, D Example Emergency Action Levels: Emergency Action Levels (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 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 failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. Guideline exposure levels beyond the site boundary.

Page 63 of 66

NEI 99-01 Rev 6 ]Proposed EAL j Justification HA6 HA7 Initiating Condition: ALERT Initiating Condition: H No Change U Difference 1 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:

AlU 1, 2, 3, 4, 5, 6, D Example Emergency Action Levels: Emergency Action Levels (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 an actual or potential that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that 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 involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. limited to small fractions of the EPA Protective Action Guideline exposure levels.

Page 64 of 66

NEI 99-01 Rev 6 Proposed EAL [Justification HU7 W-1 HU7 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change 1 Difference FIDeviation Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT. Other conditions existing which in thejudgment of the Emergency director warrant declaration of an UNUSUAL EVENT.

Operating Mode Applicability:

AlI 1,2,3,4,5,6, D Example Emergency Action Levels:

Other conditions exist which in the judgment of the Emergency Director indicate Emergency Action Levels (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 Proposed EAL Justification E-HU1m Initiating Condition: UNUSUAL EVENT [-j No Change [ Difference [j Deviation Damage to a loaded cask CONFINEMENT BOUNDARY. TMI Station does not have an ISFSI, Operating Mode Applicability:

All Example Emergency Action Levels:

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

Page 66 of 66

Tkrgm U11m lalonA Qfatinn Anndmv PI:Yslnn h,:l"Aar Tkra~ u;ia I@IavmI ~*a*ij~mi Auinww Fv~Ir~n FJm.rI~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARG1 Initiating Condition:

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

Operating Mode Applicability:

1,2,3, 4,5,6, D Emergency Action Level (EAL):

Notes:

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

e 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 folWIG ng radiation monitor-s greater than the reading shown for 15 minutes Or longer:

(cito SPoific moniRtor list and threshold v:alues)

1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 minutes.

OR

2. Dose assessment Using actual meteorology indicates doses at or beyond (site-

.... ;f*c doco r..ec.. Point) the site boundary of EITHER:

a. > 1000 mRem TEDE OR

b. > 5000 mRem CDE Thyroid OR Month 20XX TMI 3-1 EP-AA-1009 (Revision XX)

hroa Uila hainnel qtntinn Annoy I=ynlnn N.nlpar 1 IIMii isand II VtsatI AnnII IIE vl inn NucleaIr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS r-:iJ -

ReluS~ve rA,.ArU4A r.,%iy8+- a.9 Q teFbeetts

/%1 .... I * . ..'-- -- II.... J ...... JL__ __ - - -__ - _ aL_ aa r'/'#',

l ,L _ -- A A .JA -- -- A* !A.... L---- P Closed windew dese FaTes WeffieF MaR JWWU FAPVRF eXPeGleO W GGRIIRUe f9f 60rMAtesn OrIrr(jnr I I ia I II ; I

P for one hour Of inhalation.

3. Field survey results at or beyond the site boundarv 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.

Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency RM-G-25 (Cond Offqas) 9.53 E+06 mR/hr RM-A-8GH (Station Vent) 3.09 E+05 cpm RM-G-24 (RB Purge) 5.55 E+05 mR/hr RM-A-14 (ESF Vent) 6.66 E+02 uCi/cc 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 mR-em while the 5000 mR-em thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

larssjification based on effluent monitor readings assumos that a release path to th enVironm~ent is, established. if the effluent flow past an effluont moniorF 06 known to hav Month 20XX TMI 3-2 EP-AA-1 009 (Revision XX)

hrom Uila lainnei Qtatinn Annov FYAInn N.*l*ar

~tatin Anny Thr~Mu. chani Fvinn Nucler~I TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS stopped due to actioea to- ioel-ate the releaso path, then the offluent moni;tor reading is-no longor valid for clascification purpococ.

Basis Reference(s):

1. NE 199-01 Rev 6, AG1

2. EP-EAL-0609 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values, Three Mile Island

3. EP-AA-1 12-500 Emergency Environmental Monitoring

4. FSAR Section 11.4 Radiation Monitoring System

5. EP-AA-1 10-200 Dose Assessment

6. EP-AA-1 10-201 On Shift Dose Assessment Month 20XX TMI 3-3 EP-AA-1009 (Revision XX)

Three Mile tIslnd Station Annex E~xelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARSl Initiating Condition:

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

Operating Mode Applicability:

1,2,3,4,5,6,D Emergency Action Level (EAL):

Notes:

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

e 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.

e 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 mon.itors greater than the reading shown for 15 minute or O lenger:

(cite*p*ePifi; moRnitor list and threshold values)

(2) Doeo assessment using acatual mneteorology indicates, doses greater than 100 nre~m TEDES or 50 mrom thyroid CDE at Or boyen d (cite-specific dose recepto (3) Field cur.'ey rocults indicate EITHER of the following at Or beyond (cite specific doe receptor pGOit):

" Closed window dose rates greater- than 100 mR/hr expected to continue for 60 m~inutes

" Analyses of field sun~ey samples indieate thyfeid CDE greater- than 500 maemn for: one hour-of inhalationi.

1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 minutes.

OR Month 20XX TMI 3-4 EP-AA-1009 (Revision XX)

Throp RAila 1-cland Atntinn Annpv I:*lnn N.nlpnr Thre Mil l~IanrI~tetnn AnexExelon Nucl~ear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. Dose assessment usina actual meteoroloav indicates doses at or bevond 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 boundarv indicate EITHER:

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

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

Table R1 Effluent Monitor Thresholds Effluent Monitor Site Area Emergency RM-G-25 (Cond Offgas) 9.53 E+05 mR/hr RM-A-8GH (Station Vent) 3.09 E+04 cpm RM-G-24 (RB Purge) 5.55 E+04 mR/hr RM-A-14 (ESF Vent) 6.66 E+01 uCi/cc 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 mR-em while the 500 mR-em thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

%*~!1!k! L....J -- .EtI....L .!J. .'* -A AI AJ -- I-l G-asiairatcaion hased on v.. crnuten Imon-Tior r*SoAuinu assumos ;Rna

a r eicase eatrn to Month 20XX TMI 3-5 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS the environmcnt is octablished. if the cfflucnt flow past an effluent.mntri known to have stopped due to actions to ilthe thoreease path, then the effluent monGitor readg is no longer valid for claesificatio pupoos Escalation of the emergency classification level would be via IC RAG1.

Basis Reference(s):

1. NEI 99-01 Rev 6, AS1

2. EP-EAL-0609 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values, Three Mile Island

3. EP-AA-1 12-500 Emergency Environmental Monitoring

4. FSAR Section 11.4 Radiation Monitoring System

5. EP-AA-1 10-200 Dose Assessment

6. EP-AA-1 10-201 On Shift Dose Assessment Month 20XX TMI 3-6 EP-AA-1 009 (Revision XX)

ExAIon NuelAar Thriat Mia Ihqanrd Station Annex ExeInn Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARA1 Initiating Condition:

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

Operating Mode Applicability:

1,2,3, 4,5,6, D Emergency Action Level (EAL):

Notes:

9 The Emergency Director should declare the Alet-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.

e 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.

for 15 m"inutes or longer:

(site6pecific moRntor list and threshold v.'alues)

(2) Dose assessment using actual metoorolOg'; indicates dosees greater than 10 mrcm TEDE or 50 mromn thyroid CDE at or beYOnd (site spocific dose recee9tor pEGinfl. l i *

(3) Analysis of a liquid effluent sample indicates a concentration Or release rate that would result indosces greater than 10 mrom~ TEDE Or 50 mromn thyroid CDE ato beyond (site-specific dose receptor point) for ono hour Of exposure-.

(4) Field su...ey results indicate EITHE.R o.f t.he following at or beyond .(sit, pecific dose receptor pint):

4* *H

Closed window dose rates ereater- than L0 fo'hr exoeetd to eantinue fer 60 fmutes or lenger-.

a ¶ Jl" * *1 " ] f"*T'*T" a~~~ v..%= a*'" ar'~~ aafia + at-Pr fa"= mrr AW 41MJ4 ARP v a-*"~ a~

'-*-U*"

LAAA.

j b- -

hour of inhalation.

Month 20XX TMI 3-7 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

1. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 15 minutes.

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.

Table R1 Effluent Monitor Thresholds Effluent Monitor Alert RM-G-25 (Cond Offqas) 9.53 E+04 mR/hr RM-A-8GH (Station Vent) 3.09 E+03 cpm RM-G-24 (RB Purge) 5.55 E+03 mR/hr RM-A-14 (ESF Vent) 6.66 E+00 uCVcc Month 20XX TMI 3-8 EP-AA-1 009 (Revision XX)

Thrgna MiIA lalanfi Qtafirnn Annov IFwalnn N, Aalar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 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 mR-em thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classiffication based on effluent monefitonr re-adfings acsumos that a roleaso path to the enViroAnment is established. If tho effluent flow-past. an Aeffiluent moenitor is known to have stopped duo to acti;ons to isolato the release path, then the offluePnt mnonitor reRaing

s no Ilnger valid forclasif;cation purpoes.7 Escalation of the emergency classification level would be via IC RAS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA1

2. OP 1101-2.1 Radiation Monitoring System Setpoints

3. FSAR Section 11.4 Radiation Monitoring System

4. OP-TM-MAP-CO101, Radiation Level HI

5. EP-EAL-0609 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values, Three Mile Island

6. EP-EAL-0616, Revision 0, Three Mile Island Criteria for Choosing Radiological Liquid Effluent EAL Threshold Values

7. EP-EAL-0622, Revision 0, Three Mile Island Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values for Waste Gas Decay Tanks Month 20XX TMI 3-9 EP-AA-1 009 (Revision XX)

Tkrg%,A% Uilg% lalonri Qfation Annav I:valrnn Hnn-lmar Tkr~

BB 1 MiI~

W*BB I.IanrI

5* 5 Qt~tirui tt*WU . Ann~v rB. B. Fv.Inn j,~ * *Muu~I.~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS ARUM Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM (kste-speelfiG efflu.nt

. eloase controlling document) limits for 60 minutes or longer.

Operating Mode Applicability:

1,2, 3, 4, 5, 6, D Emergency Action Level (EAL):

Notes:

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

(1) Reading on ANY effluent radiation monitor gr*eater than 22timnes t-he- (site specific effluent release oRntrolling d.,umont). Imits for 60 minutes Or Ilnger (cite-specific monitor list and threshold values corresponding to 2 times the Controlling dom*ent* I-Mits (2) Reading on ANY effluent radiation monitor greater than 2 times the alarm setp established by a current radioactiveity discharige permit for 60 minutes or longer.

(3) Sample analysis for a gaseous Or liquid release indicateS -aconcentration or release rate greater than 2 times th (site-specfific effluent release controllig documnent) limnits for 60 minutes 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.

i. RM-L-6, Radwaste Discharge ii. RM-L-1 2, IWTS / IWFS Discharge iii. RM-A-7, Waste Gas Decay Tank Discharge iv. Discharge Permit specified monitor OR

2. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 60 minutes.

OR Month 20XX TMI 3-10 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS

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.

Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Event RM-G-25 (Cond Offoas) 1.09 E+03 mR/hr RM-A-8GH (Station Vent) 7.03 E+01 cpm RM-G-24 (RB Purge) 6.34 E+01 mR/hr RM-A-14 (ESE Vent) 7.60 E-02 uCi/cc 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.

C-larssific-attion based on effluent moenitor ro-ad-ings assumes that a rclease patht the enVrOFnment ir, established. if the effluent flow p'tanefluent moni~tor is knownt have Mtopped duo to ac-tions to islthe- Ih then the effluent monitor road-ing

'eosah, sc no longer valid for clIiiaioGupss 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 oathwavs (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 Month 20XX TMI 3-11 EP-AA-1009 (Revision XX)

hrgga Uffis lalzanA Qtatinn Anniov i=Yplnn NiJnlpnr Thr~ MIle Iclenri ~t~atinn Annev Fv~Inn Niir~Ic&~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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-#1---This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous erI-iquid-effluent pathways.

EAL #3 Basis:

EAL--#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. OP 1101-2.1 Radiation Monitoring System Setpoints

3. FSAR Section 11.4 Radiation Monitoring System

4. Offsite Dose Calculation (ODCM)

5. OP-TM-MAP-CO101, Radiation Level HI

6. EP-EAL-0609 Revision 1, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values, Three Mile Island Month 20XX TMI 3-12 EP-AA-1 009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclanr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA2 Initiating Condition:

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

Operating Mode Applicability:

1,2,3, 4,5,6, D Emergency Action Level (EAL):

(1) Uncovor'; Of irradiatcd fuel in the REFUELING P.ATHWIA.IY.

2. Damage to irradiaced fuel resulting in a release of radioacivity from the fuel as indicated by ANY of R2 Radiation Monitorr i (sito Spocific listiRg of radiation monitors, and tho associated roadings, s Rtpoints and/RB alarms)

(3) LIwonng of spent fuel pool level to tu(siteepific Level 2 value). [tqh whv3ich NOtefe 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 R2 Radiation Monitor reading >1000 mRemlhr Table R2

____________ Radiation Monitors_____

RMS Area Monitored Mode RM-G-9 FHB3 Bridge Rad Monitor ALL RM-G-6 RB Auxiliary Bridge Rad Monitor 5.6 RM-G-7 RB Main Bridge Rad Monitor 5.6 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.

Month 20XX TMI 3-13 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS IMMINENT: The traiectory 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.

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly_., or a -nifclant low..ring of ,.,atar , le. within the

,pent fuel pool (See Developer.

,te,). 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 ICapplies to irradi-atteed fueel that is licensod for dr; storage up to the point ta tho loaded storage cask is soaled. Once sealed, damage to a leaded cask causing lesc of the CONFINEMENT B1OUNDARY is, classified in accrdance with CG EHLHJl.

. o tho oM

.c.alation dd bo hb.c.d on pithor

. e.ogniio.n

. o.a.... .A or C .C.

EAL #1 Basis:

E-AL-41 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 aA iieaeerise 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).

Month 20XX TMI 3-14 EP-AA-1009 (Revision XX)

I::xelon Nucl@ar Throp Mile~ Island Staition Annex Exelon Nuceianr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Escalation of the emergency would be based on either Recognition Category R A-or C EA #3 Spent fuel pool water level at this value ir, within the low,,er And of the level range neccary to prevent significant dose conseguoncos from direct gamma radiation to perconnel performing eperations in the V.icinity Of the cpent fuel peel. This condito reflects a significant loss Of spcnt fuol pool water inventor; and thus it.is alsoRa a precursor two t .. *fVtt 7 w ra ,- c5 7 5m *q5 -* W - - s - ,^ -

- !C-- - t- -- ,

Esclatonof the emorgnonc Gla6ification leveXl wo.AUld be via I~s A8i Or A82 f(se AS D v.,l,.rr Aot,).

Basis Reference(s):

1. NEI 99-01 Rev 6, AA2

2. Operating Procedure OP-TM-MAP-C0105 RCS Draindown LVL HI/LO

3. OP-TM-MAP-CO101, Radiation Level HI

4. UFSAR, Section 14.2.2.1 - "Fuel Handling Accident" 5 Technical Specification 3.8.11 (Reactor Cavity Level)

6. Operating Procedure OP 1101-2.1 Radiation Monitoring System Setpoints Month 20XX TMI 3-15 EP-AA-1009 (Revision XX)

I=*Alnn N.cIAar Thro*wisw kthand Stamtion Annexv EvAnn Nucl*war TABLE TMI 3-2: EAL Technical Basis 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, 6, D Emergency Action Level (EAL):

/i4 x I IKlDI AKllII l r-wlA in=,nI r*Irj ;r%+km OCh lII C I l In-_ DAT1IfAIAV

\'/

k.~. AKN -; 6"-n n~n.,.,.

i iýi i i F iz a (Site G~ecific loeve'l indicatiens). I AkIN

.A.ND-

-1 1- - A as%* 9 1-u= ~r.I fow S *I 1~ U U* III L4* L4 S L4U~tI~5 U

Y~U ~A USSASSA~ flSU U 5 ti 5 feiwn Faife lnenfter *

(sitospocific list of aroa radi-ation monitor)

1. a. UNPLANNED water level drop in the REFUELING PATHWAY.

AND

b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R2.

Table R2 Radiation Monitors RMS Area Monitored Mode FHB Bridge Rad Monitor ALL RM-G-9 RM-G-6 RB Auxiliary Bridge Rad Monitor 5,6 RM-G-7 RB Main Bridge Rad Monitor 5, 6 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.

Month 20XX TMI 3-16 EP-AA-1009 (Revision XX)

I=x*lon Nuclear Three Mile Island Staition Annex Exelnn Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS 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 deGrease-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.

I A water level deGrease-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 inRGeaserise 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 iGFreaserise 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. UFSAR, Section 14.2.2.1 - "Fuel Handling Accident"

3. OP-TM-MAP-CO105 RCS Draindown

4. OP 1202-12, Excessive Radiation Levels

5. OP 1101-2.1 Radiation Monitoring System Setpoints

6. Technical Specification 3.8.11 (Reactor Cavity Level)

Month 20XX TMI 3-17 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS ARA3 Initiating Condition:

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

Operating Mode Applicability:

1,2,3,4,5,6, D Emergency Action Level (EAL):

Note: If the equipment in the listed-room or area listed in Table R4 was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted (1) Dooe Fate gr.ator than, 15 R'h*hr in ANY of the f*llowing ar.as.:

Control Room w

9 (other-site speeitie afeas~r-eefn) fO~t An~ UKINIDIAWNIRfl Aumth mra.u, Or, *n mlotrr' I J IV T ViV that levels 5i I* anrihibet VI Vi IIVIt VI irneedc OF Ill IVV*V KVVVVV t aGcrns tV 1 4 a ~ S I.~

1I a II a..a.r.n rI y. .h* main a. am flmaa% as.

__. J - - - -- I--

vv IEI I I R I i isitelsoeGGisi t or plant rooms orareas wutn enir: roiatoa moac avvilcaHl)lty l

1. Dose rate QreateF than> 15 mR/hr in ANY of the areas contained in Table R3:

Table R3 Areas Requiring Continuous Occupancy

Main Control Room

Central Alarm Station - (by survey)

OR Month 20XX TMI 3-18 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Exelon Nuclear Nuclear Three Mile Island Station Annex TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to any of the following Table R4 plant rooms or areas:

Table R4 Areas with Entrv Related Mode Aplicability Area Entry Related Mode Applicability Reactor Building* Modes 4, 5. and 6 Intermediate Building* Modes 4. 5. and 6 Auxiliary Building* Modes 4. 5, and 6 Fuel Handling Bujildng* Modes 4. 5. and 6

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 procedures--maintain normal plant .p..atin, Or to pcftm a normal plant cooldown and sh'Jtdo-n. 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 R4 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which reguire 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 reguired 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 tho tome of the clevated radiation le:ecl. 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 Month 20XX TMI 3-19 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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).

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 iP,-easerise 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 BA, C or F ICs.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA3

2. FSAR Section 5.01 Class I Structures, Components, and Systems

3. OP-TM-MAP-CO101, Radiation Level HI Month 20XX TMI 3-20 EP-AA-1009 (Revision XX)

Throm Milo lalanrl Qtafinn Annoy Fyislnn Nurnlpar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS SRU3 Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability:

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

ran a.l .n a a ran*n.. *k nn ia~v,

_*-* ^;;* S,,l1, ,.-*

I - ~

'-I - ....... * ..... r

~ 3~I~I mni IfOICTO~Tf3T rnc~o colan aciv~y .....- -..- j--

.mio i

oraTo ma a

il k! !; ;+ of I A I -r k ; i Q ;f; +; n Ct %7vrct W M OPWID CY ri V0 ri out F7%70 %7= 17 17-

1. Letdown Monitor RM-L-1 alert alarm (high or low channel).

OR

2. Sample analysis indicates that:

a. Dose Equivalent 1-131 specific coolant activity > 60.0 uCVgm.

OR

b. Dose Equivalent XE-133 specific coolant activity> 797.0 uCi/lm.

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 BA ICs.

Month 20XX TMI 3-21 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Basis Reference(s):

1. NEI 99-01 Rev 6, SU3

2. Operating Procedure 1101-2.1 Radiation Monitoring System Setpoints

3. FSAR Section 11.4.4, Liquid Monitoring Subsystem

4. OP-TM-MAP-C0101, Radiation Level HI

5. Technical Specifications 3.1.4, Reactor Coolant System Activity Month 20XX TMI 3-22 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis 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:

1,2,3,4 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-F-3 Month 20XX TMI 3-23 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:

Loss or Potential Loss of ANY two barriers.

Operating Mode Applicability:

1,2,3,4 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-3 Month 20XX TMI 3-24 EP-AA-1009 (Revision XX)

Thrao M~iId Wcandl Qtatinn Annoy I=:talnn Nhlo irlr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:

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

Operating Mode Applicability:

1,2,3,4 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-3 Month 20XX TMI 3-25 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:

RCS or SG Tube Leakage Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

POTENTIAL LOSS A. RCSloactor v~eseol leveol less than (site spccific value).

1. RCITS hot leg instruments indicate 0 inches after lowering trend.

AND

2. In-core thermocouples are unavailable.

AND

3. ALL RCP's are secured.

Basis:

There is no Loss threshold associated with ROS or SG Tube Leakage.

Potential Loss Threshold Basis:

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. FSAR 7.3.2.2.c. 10.d

3. OP-TM-EOP-008, RCS Superheated

4. OP-TM-EOP-010, Emergency Procedure Rules Guides and Graphs Month 20XX TMI 3-26 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:

Inadequate Heat Removal Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS A. Core Exit Them~ecouplo readings greater than (site sepoific temper-ature value)

1. TcŽad > 1400°F Potential Loss A.Co re exit thorMOcouplo reading s greater than (site sepocfic temperature value).

n~. - ^S- - - - - -aIn .... .% d "ry ry k;!;#I~~t .V f rt n4%nv

r. mrtr" ""m

-4%San,%e "M -Mft-

~ A;lt*i 11"U."U" V5 A k~ (.4r Ift "

-- r -- ,# * ....

cpocific niain)

2. > 25°F Superheat OR

3. HPI-PORV Cooling in effect Basis:

Loss Threshold #1 Basis L~es&-2-A ThisFeadmig-ilndicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

Potential Loss Threshold #2 Basis Potential Loes 2.A.

ThiSAd-rm*-*_lndicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.

Potential Loss Threshold #3 Basis Potentia*l L ,,os [2B This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the Month 20XX TMI 3-27 EP-AA-1009 (Revision XX)

Nuclear I=x*lon Nuclasr Three Mile Island Station Annex Exellon TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

Meeting this threshold results in a Site Area Emergency because this threshold is I identical to RCS Barrier RC2 Potential Loss threshold-2-A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat I removal may result in fuel heat-up sufficient to damage the cladding and inireaseraise RCS pressure to the point where mass will be lost from the system.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. OS-24 Attachment D

3. OP-TM-EOP-004, Lack of Primary to Secondary Heat Transfer

4. OP-TM-EOP-008, RCS Superheated

5. OP-TM-EOP-010, Emergency Procedure Rules, Guides And Graphs

6. OS-24, Conduct of Operations during Abnormal and Emergency Events Month 20XX TMI 3-28 EP-AA-1009 (Revision XX)

Thrdam Uilm lainnel tatinn Annoy I=Yplnn Nnnlanr Tkraa Mum IcImnrI ~tmtinn Ann.v Fv.Inn NuIr~Isr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC3 Initiating Condition:

Containment Radiation / RCS Activity Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS 1A. Containment radiation monitor roading grcatc. than *(site epcific valuc)

(RM-G-22 or RM-G-23) reading > 1.95E+03 R/hr.

OR

?8. (Sitoe-pocific indications that roactOr coolant activ.ity is groator than 300uGi'gm dose equivalent 1131) Coolant activity as sampled > 300uCi/gm Dose Equivalent 1-131 Basis:

Loss Threshold #1 Basis Less4-3A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300jiCi/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 RC3 Loss Tthreshold &A-since 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.

Loss Threshold #2 Basis This threshold indicates that RCS radioactivity concentration is greater than 300 RCi/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 Month 20XX TMI 3-29 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION There is no Fuel Clad Barrier Potential Loss threshold associated with RCS Activity /

Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. OP-TM-MAP-CO101, Radiation Level HI

3. FSAR Section 11.4.4, Liquid Monitoring System Description

4. Calculation C3640-98-034, Prediction of the Response of RM-G-6 and 7 to Fuel Damage

5. Core Damage Assessment Methodology (CDAM)

Month 20XX TMI 3-30 EP-AA-1 009 (Revision XX)

hrAmAm Uila lalonrl Qtafinn Annov F:yalnn N"~lAar Tkra~ Mum I@kanrl ~t~tinn Anninv Fvmlnn M.mr~I~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC56 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS 1_A. 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 This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.

Potential Loss Threshold #2 Basis Potential Los 6. A.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 9-F-3 Month 20XX TMI 3-31 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Exelon Nuclear Nuclear Three Mile Island Station Annex TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC1 Initiating Condition:

RCS or SG Tube Leakage Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS A-.1 .AA Aautomatic or manual EGGS -(S0)ESAS actuation is required by EITHER of the following:

ag. UNISOLABLE RCS leakage OR

!22. Steam Generator tube RUPTURE.

POTENTIAL LOSS

2. UNISOLABLE RCS leakage > 150,qpm.

OR

3. HPI-PORV Cooling in effect.

OR

4. a. RCS Pressure > 2450 psic.

AND

b. RCS Pressure not lowering.

OQperation of a standby charging (makeup) pum Ncrqired by EITHER ofth following.-!. UIISOLA\B*ER*C

[- S leakage SG tube leakage.

B1. . RCS coldWnr rFate greater than (site pe,;ifi, pres*rized thermal -hock cicra'imtcdefined by ciGseii indicatienc Month 20XX TMI 3-32 EP-AA-1009 (Revision XX)

Throa MUil WaIndl Qtztinn Annoiy Fyalnn Nnnlanr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION mm * *

li'A"

3i. I-". !ntO::ltV .I.too entr I conat.... mot Basis:

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

RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety iniection.

FAULTED: The term aDDlied to a steam aenerator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Loss Threshold #1 Basis Losse-1A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier CT1 Loss threshold 1-A-will also be met.

Potential Loss Threshold #2 Basis Potential Lss 1 .A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.

Month 20XX TMI 3-33 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold CT1 .A-will also be met.

Potential Loss Threshold #3 Basis Patontial I o-.c 1R This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. OP-TM-EOP-010 Emergency Procedure Rules, Guides And Graphs

3. OP-TM-EOP-002 Loss of 25 0 F Subcooled Margin

4. OP-TM-MAP-D031, MU Flow HI

5. OP-TM-EOP-005, OTSG Tube Leakage Month 20XX TMI 3-34 EP-AA-1 009 (Revision XX)

Three Mile Island Station Annex Exelon Exelon Nuclear Nuclear Three Mile Island Station Annex TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:

Inadequate Heat Removal Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

POTENTIAL LOSS A.

HPI-PORV Cooling in effect inadequate ROS ho-at .mo.a. ,iacapability stoam g.n.rato.. as indicated by i;,

cpociftcidctoc Basis:

There is no Loss threshold associated with Inadequate Heat Removal.

Potential Loss Threshold Basis Potential Loss 2.-A HPI-PORV Cooling in effect indicates a Lack of Primary to Secondary Heat Transfer capability.

This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier FC2 Potential Loss Tthreshold 2-#8#3; both will be met.

This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and i,,reaseraise RCS pressure to the point where mass will be lost from the system.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. OP-TM-EOP-004, Lack of Primary to Secondary Heat Transfer Month 20XX TMI 3-35 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:

Containment Radiation / RCS Activity Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS A. Containment radiation monitor reading grato, than (RM-G-22 or RM-G-23) reading >

25 R/hr.

-.(site specific value).

Basis:

Loss Threshold Basis The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier FC3 Loss Tthreshold 3-.A#1 since it indicates a loss of the RCS Barrier only.

I There is no RCS Potential Loss threshold associated with RCS Activity / Containment Radiation.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. EP-EAL-061 1, Criteria for Choosing Containment Radiation Monitor Reading Indicative of Loss of RCS Barrier Month 20XX TMI 3-36 EP-AA-1009 (Revision XX)

IFvnlnn Nudla~r Thrais Milp Island Station Annax Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC56 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS I1_A. Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.

POTENTIAL LOSS 2B. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.

Basis:

Loss Threshold #1 Basis Lsse-&A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.

Potential Loss Threshold #2 Basis P*tential Loe* 6. A-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 9-F-3 Month 20XX TMI 3-37 EP-AA-1 009 (Revision XX)

Throm Uila lainnei Qtatinn Annoy F:Yalnn N, nrlanr Turn, MiI~ It6nrl ~tmtinn Ann.v Fv.Inn Mmu~I~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT1 Initiating Condition:

RCS or SG Tube Leakage Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS A. A leaking or RUPTURED SG is FAULTED outside of containment.

1. SG tube leakage > 150prm AND

2. SG is FAULTED outside of containment.

Basis:

RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety iniection.

FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Loss Threshold Basis This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier RC1 Potential Loss Threshold 2.b 4-A and Loss Threshold 1.b-A, respectively. This condition represents a bypass of the containment barrier.

FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC Month 20XX TMI 3-38 EP-AA-1009 (Revision XX)

Ex*lon Nuclear Thropu MiIA 1-cland Atnatinn Annex Exelnn Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION BRSU3 for the fuel clad barrier (i.e., RCS activity values) and IC MSU64 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve)-dG meets this threshold.

Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category BA lCs.

The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Month 20XX TMI 3-39 EP-AA-1009 (Revision XX)

Throm Milo Ilainel ~tatinn Annoyv F:yalnn Nwlona~r TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Affected SG is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 25 No classification No classification gpm (Or othor value p..

Grete thanm 25 gpm (o SUW Do'v.'lepor Notes)

Greater than 25 gpm(o Unusual Event per Unusual Event per Oth'r v.,alue p*r SU4 SU4MU6 SU4MU6 Greater than 150 aDm.

The capacity of one makeup pump in the normal Site Area Emergency charging mode is exceeded per FS1 Alert per FA1 Requires operation of-a pumpo(RCS Barrier PotentialLoss)

Requires an automatic or Site Area Emergency Alert per FA1 manual EGGS (SI) actuation per FS1 (RCS BarrierLoss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. OP-TM-EOP-01 0, Emergency Procedure Rules, Guides And Graphs

3. OP-TM-EOP-005, OTSG Tube Leakage

4. OP-TM-EOP-001, Reactor Trip Month 20XX TMI 3-40 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exellon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:

Inadequate Heat Removal Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

POTENTIAL LOSS A. 1. (Site cpocific critercia forF Antry into core cooling roctoration procoduro)

AND

2. RestorFation pmroedure not eftoctiI.oA Within 15 mninue

1. Tcad > 1800°F AND

2. EOP Restoration procedures not effective in < 15 minutes.

Basis:

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.

There is no Loss threshold associated with Inadequate Heat Removal.

Potential Loss Threshold Basis Potential Loes 2.A.

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

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

Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Month 20XX TMI 3-41 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. OP-TM-EOP-008, RCS Superheated

3. OP-TM-EOP-010, Emergency Procedure Rules, Guides And Graphs

4. OP-TM-EOP-006, LOCA Cooldown Month 20XX TMI 3-42 EP-AA-1 009 (Revision XX)

E~xAlnn Nudlanr Throgm RAilas Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Containment Radiation / RCS Activity Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

POTENTIAL LOSS AX Containment radiation monitor roading gr.ator than (RM-G-22 or RM-G-23) reading >

4.40E+03 R/hr.(scite sp.cifc value).

Basis:

There is no Loss threshold associated with RCS Activity / Containment Radiation.

Potential Loss Threshold Basis Potontial Loss 3. A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the 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 Nuclear Power 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-3

2. Core Damage Assessment Methodology (CDAM)

Month 20XX TMI 3-43 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Exelon Nuclear Nuclear Three Mile Island Station Annex TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT4 Initiating Condition:

Containment Integrity or Bypass Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS 1A. Containment isolation is required AND EITHER of the following:

a4-. UNPLANNED deGrease-lowering in containment pressure or rise in radiation monitor readings outside of containment in the Emergency Directors judgment that indicate a loss of containment integrity.

OR b2. UNISOLABLE pathway from containment to the environment exists.

OR

28. Indication of RCS leakage outside of containment POTENTIAL LOSS JA nmnerI-O nIY conaiiionS menieacior BuilOing pressure > 5o pslg and rising.

OR

48. Explosive mnixture exists inside containment.. Hydrogen Concentration in Containment > 4%.

OR 5G. al-. GentainmeFt-Reactor Building pressure greater than (site spe.ifi; pressure setp t> 30 osici AND b2. ILees~ than one full per deign for 15 ute train of (site sepoific. Syrtcm Or equipment) is operating "min,or longer. Reactor Building Emerqency Cooling is less than ANY one of the following conditions.

SPRAY COOLERS 2 0 0 3 1 1 Basis:

FAULTED: The term aDDlied to a steam aenerator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Month 20XX TMI 3-44 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION 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.

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

LossThreshold #1 Basis:

L-ess-4.

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both loss thresholds 1.a4--A4 and 1.b4.A2.

1.a4 A I - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g.,

containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

Refer to the middle piping run of Figure 9-F-4. Two simplified examples are provided.

One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category RA ICs.

1.b4*A Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere Month 20XX TMI 3-45 EP-AA-1009 (Revision XX)

Thrno Wig%Ilainni Qtafinn Annov Fvalnn Nwlon*ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION (e.g., through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 9-F-4. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,

containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.

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.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 9-F-4. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met.

If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then loss threshold 2_4B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 1 .a4A4 to be met as well.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category RA ICs.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Barrier CT1 Loss lthreshold-l-A.

Loss Threshold #2 Basis:

L4e66-4.B Containment sump, temperature, pressure and/or radiation levels will inGoeaserise if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). IFiGeaseRaises in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.

Month 20XX TMI 3-46 EP-AA-1 009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not inareaserise significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.

Refer to the middle piping run of Figure 9-F-4. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building.

Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause loss threshold 1 .a4.A-.- to be met as well.

To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Barrier RCl Loss Threshold 1.a and/or Potential Loss threshold 2.a-l-A to be met.

Potential Loss Threshold #3 Basis PtentiaI Less 4. A If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.

Potential Loss Threshold #4 Basis Potential Loss 4.1.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Potential Loss Threshold #5 Basis Potontial L-se 4.G This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc.,

Month 20XX TMI 3-47 EP-AA-1009 (Revision XX)

Throo Milo Icinnel Qtnfinn Annov F:valnn Iw-la~l r Thr~ Miki IQI~nrI ~t~tinn Ann~v Fv~Inn Miir~I~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION but not including containment venting strategies) are either lost or performing in a degraded manner Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-3

2. FSAR Section 6.6 Reactor Building Pressure-Time Response

3. Technical Specifications 3.5.3, Engineered Safeguards Protection System Actuation Setpoints

4. FSAR Section 6.3.3, Actuation

5. FSAR Section 6.5.3, Hydrogen Generation

6. OP-TM-EOP-006, LOCA Cooldown

7. 1302-5.25 Reactor Building Sump Level

8. FSAR Section 5.2 Reactor Building

9. FSAR Section 9.2, Chemical Addition and Sampling System

10. OP-TM-EOP-002, Loss of 250 F Subcooling Margin

11. OP-TM-EOP-006, LOCA Cooldown

12. OP-TM-EOP-010, Emergency Procedure Rules, Guides and Graphs

13. OP-TM-MAP-D0301, High Make-up Flow Month 20XX TMI 3-48 EP-AA-1009 (Revision XX)

Tknagi RAHft Id.-AftrvA Q*-m*monU I~EEE Anniaw Ivaletn Mmrai lnar IEEE ~ EVIIE~ U~E~E E~4 ~ E~*~ ~~~~U~SU U***

TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT56 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

1,2,3,4 Fission Product Barrier (FPB) Threshold:

LOSS

.1A. Any condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.

POTENTIAL LOSS

28. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.

Basis:

Loss Threshold #1 Basis LOesS-6 This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

Potential Loss Threshold #2 Basis Potential Loss 6.,A 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, Table 9-F-3 Month 20XX TMI 3-50 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSG1 Initiating Condition:

Prolonged loss of all Off-site and all On-Site AC power to emergency busses.

Operating Mode Applicability:

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

Note: The Emergency Director should declare the General Emorgec-*--.vent promptly upon determining that (cite cp-.ifi," hou .r,) the applicable time has been exceeded, or will likely be exceeded.

1--a-. Loss of ALL offsite and ALL , n"it; AC power to (Sit .specific em*.

egency buses)Emeraencv 4KV buses.

AND

2. Failure of EG-Y-1 A, EG-Y-11B Emerqency Diesel Generators and EG-Y-4 SBO

. Diesel Generator to supply power to Emergency 4KV buses.

AND 3b. EITHER of the following:

a. Restoration of at least one emer4eRey-Emergency 4KV bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is noil*ss than (site specific hourS) i. n likely.

OR

b. > 25 0F superheat p

I *Ill m I I I t~ite sE~ocmc inaication oran ina~iiitv toaaoeauateiv remove noa;t mrnrom

l o orc J Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or lacin it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

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 removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of oRe rOmoroa ny fission product Month 20XX TMI 3-51 EP-AA-1009 (Revision XX)

hrom Wha lalanri tatinn Annov F::alnn Kliinl*ar Tkrn~ MiI~ Iokanrl ~hatinn Ann.v FvInn fJmur~I.~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 emergency busses of the affected unit can be powered from the unaffected unit through the crosstie breakers. Unit crosstie is considered an adequate source of offsite power when evaluating this EAL.

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. OP-TM-EOP-01 0 Emergency Procedure Rules, Guides And Graphs

3. FSAR Section 8.2.2 Unit Distribution System

4. FSAR Section 8.2.3 Sources of Auxiliary Power

5. FSAR Section 8.5 Station Blackout

6. Technical Specification Section 3.7, Unit Electric Power System

7. 1107-1 Normal Electrical System

8. 1107-2A Emergency Electrical - 4KV and 480 Volt

9. OP-TM-AOP-020 Loss Of Station Power

10. 1107-3 Diesel Generator

11. 1107-9 SBO Diesel Generator

12. OP-TM-EOP-008 RCS Superheated

13. OS-24, Conduct of Operation during Abnormal and Emergency Events Month 20XX TMI 3-52 EP-AA-1 009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSS1 Initiating Condition:

Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer.

Operating Mode Applicability:

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

Note: The Emergency Director should declare the Sito Arop E..ergencyevent promptly upon determining that the applicable time 4-5-inutes-has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite and ALL ,n-it, AC Power to (;ito .. p.ific oFmo.gon..Y buses)Emergency 4KV buses for 15 minutec Or longR. .

AND

2. Failure of EG-Y-1A, EG-Y-11B Emergency Diesel Generators and EG-Y-4 SBO Diesel Generator to supply power to Emergency 4KV buses.

AND

3. Failure to restore power to at least one Emergency 4KV 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.

The emergency busses of the affected unit can be powered from the unaffected unit through the crosstie breakers. Unit crosstie is considered an adequate source of offsite power when evaluating this EAL.

Month 20XX TMI 3-53 EP-AA-1009 (Revision XX)

Throa Wig%lalanfi Qfafinn Anngv Fvalnn .iollar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RAG 1, FG1. e--MSG1 1 or MG2.

Basis Reference(s):

1. NEI 99-01 Rev 6, SS1

2. FSAR Section 8.2.3, Sources of Auxiliary Power

3. Technical Specification Section 3.7, Unit Electric Power System

4. 1107-1 Normal Electrical System

5. 11 07-2A Emergency Electrical - 4KV and 480 Volt

6. OP-TM-AOP-020 Loss Of Station Power

7. 1107-3 Diesel Generator

8. 1107-9 SBO Diesel Generator

9. FSAR Section 8.2.2, Unit Distribution System

10. FSAR Section 8.5, Station Blackout Month 20XX TMI 3-54 EP-AA-1 009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSA1 Initiating Condition:

I Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

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

Note: The Emergency Director should declare the eventAleFt promptly upon determining that the applicable time 16 r:inutes-has been exceeded, or will likely be exceeded.

1. AC power capability to Emergency 4KV buses reduced to only one of the following power sources for > 15 minutes.

" Auxiliary Transformer 1A

" Auxiliary Transformer 1B

Emergency Diesel Generator EG-Y-1A

" Emergency Diesel Generator EG-Y-1 B

" SBO Diesel Generator EG-Y-4

a. AC power vapability' to (cite* pc*.ific cme.e*gny buses) is reduced to a singl.

9ower source for 15 minutes Or longer.

AND 2b. Any additional single power source failure will result in a loss of all-ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A svstem reauired for safe Dlant operation, cooling down the plant and/or Dlacina it in the cold shutdown condition.I includina 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 MSUI.

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 TMI 3-55 EP-AA-1009 (Revision XX)

hrgaia Wha lalanA Qtatinn Annov P::ainn Niinlan*r Throo Milo IeIonrI Qtoti,~n Annoy Fv~Inn NmIr~I~r TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS e A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

e A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

9 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. FSAR Section 8.2.3, Sources of Auxiliary Power

3. FSAR Section 8.5, Station Blackout

4. Technical Specification Section 3.7, Unit Electric Power System

5. 1107-1 Normal Electrical System

6. 11 07-2A Emergency Electrical - 4KV and 480 Volt

7. OP-TM-AOP-020 Loss Of Station Power

8. 1107-3 Diesel Generator

9. 1107-9 SBO Diesel Generator

10. FSAR Section 8.2.2, Unit Distribution System Month 20XX TMI 3-56 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSU1 Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

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

Note: The Emergency Director should declare the Unusual E'-entevent promptly upon determining that the applicable time 1-6 Finutes-has been exceeded, or will likely be exceeded.

1-.Loss of ALL offsite AC power capability to Emergency 4KV buses (site-speGifi emcrgoncy b.. , ) for _>15 minutes-er-feng.

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.

The emergency busses of the affected unit can be powered from the unaffected unit through the crosstie breakers. Unit crosstie is considered an adequate source of offsite power when evaluating this EAL.

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 MSA1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SUW

2. FSAR Section 8.2.3, Sources of Auxiliary Power

3. Technical Specification Section 3.7, Unit Electric Power System

4. 1107-1 Normal Electrical System

5. 11 07-2A Emergency Electrical - 4KV and 480 Volt

6. OP-TM-AOP-020 Loss Of Station Power

7. 1107-3 Diesel Generator

8. 1107-9 SBO Diesel Generator Month 20XX TMI 3-57 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS

9. FSAR Section 8.2.2, Unit Distribution System Month 20XX TMI 3-58 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSG28 Initiating Condition:

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

Operating Mode Applicability:

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

Note: The Emergency Director should declare the Genoral ,, ,.,-ge..yevent promptly upon determining that the applicable time 46-eninutes-has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to Emerqency 4KV buses.

AND

2. Failure of EG-Y-1A, EG-Y-1B Emerqency Diesel Generators and EG-Y-4 SBO Diesel Generator to supply power to Emergency 4KV buses.

AND

3. Voltage is < 105 VDC on1 25 VDC Distribution System 1A and 1 B.

AND

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

1. a. Los_6ef o ALL-'-eof 6it. a:;ndi A LL onsite AG power to (eitc specific omergeFncy

- buses) for 15 mninutes or longer-.

AND

b. Indicated voltage is less than (site-specific bus voltage value) onAL (site sp;cific; Vital D- busses) for 15 minute*s r *oFngr.

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 tvpicallv systems classified as safetv-related.

Month 20XX TMI 3-59 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 beth-all EALs thleshelds-are met.

Basis Reference(s):

1. NEI 99-01 Rev 6, SG8

2. FSAR Section 8.2.2.6, 250/125 VDC System

3. 1107-2C Vital DC Electrical System

4. OP-TM-AOP-023, A DC System Failure

5. OP-TM-AOP-024, B DC System Failure

6. FSAR Section 8.2.2, Unit Distribution System

7. FSAR Section 8.2.3, Sources of Auxiliary Power

8. Technical Specification Section 3.7, Unit Electric Power System

9. 1107-1 Normal Electrical System

10. 11 07-2A Emergency Electrical - 4KV and 480 Volt

11. OP-TM-AOP-020 Loss Of Station Power

12. 1107-3 Diesel Generator

13. 1107-9 SBO Diesel Generator

14. FSAR Section 8.5, Station Blackout Month 20XX TMI 3-60 EP-AA-1009 (Revision XX)

krAmg% Rfiligh lalonA Ctotinn Anniav IPvgmlen khaot--mar lkm~

I ll WI

1U a IaI~nI*I

usu ,.QVliruil uu. B "IlIAnul. a flu vninM ll Iut.I.I..I lar*U1 TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSS28 Initiating Condition:

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

Operating Mode Applicability:

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

Note: The Emergency Director should declare the Site Aroa Em...ge..yevent promptly upon determining that the applicable time 45-inutes--has been exceeded, or will likely be exceeded.

'i*d"ated vVoltage is < 105 VDC loss than (site spocific b.us Ivoltage value)- no 125 VDC Distribution System 1A and 11B ALL (sit-sp'. .fiG " vital DG bussee4 for

>15 minutes-er-IGRleF.

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 MSG3.

Basis Reference(s):

1. NEI 99-01 Rev 6, SS8

2. FSAR Section 8.2.2.6, 250/125 VDC System

3. 1107-2C Vital DC Electrical System

4. OP-TM-AOP-023, A DC System Failure

5. OP-TM-AOP-024, B DC System Failure Month 20XX TMI 3-61 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSS36 Initiating Condition:

Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

1. Automatic or Manual Ttrip did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

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

AND

3. EITHER of the following conditions exist:

a. Tclad > 1400 0 F.

OR

b. HPI-PORV Cooling in effect

- .fl!A...

a. cakeO sDCCITI indirtcaliiun of an

-- " \ ..... r ......

inability to adequately rumuve nueat irumA

-4' -- 1 ....... j Me GOrci OR L.. / ','

1-. laiw eciiic~m l inudaiurime of aR nauniwit touueuateiv I I I Fefflve i4A--4 'rUNM I.-

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 iniection~all subs*q.unt operator a*ct..*. to mnaually shutdown the reactr* 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 Month 20XX TMI 3-62 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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.

IA reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RAG1 or FGI.

Basis Reference(s):

1. NEI 99-01 Rev 6, SS5

2. OP-TM-EOP-004, Lack of Primary-to-Secondary Heat Transfer

3. OP-TM-EOP-01 0, Emergency Procedure Rules, Guides And Graphs

4. OP-TM-EOP-001, Reactor Trip

5. 1102-4 Power Operation

6. OP-TM-641 -000- Reactor Protection System (RPS/DSS)

Month 20XX TMI 3-63 EP-AA-1 009 (Revision XX)

Throp Milo haland qtatinn Annoy I:*Alnn Nu*lAar

~tiatinMilAnnv Thre I~Iunr Feinn Nuclesar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSA36 Initiating Condition:

Automatic or manual trip 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. A.-aAutomatic OF-manual[Ttrip did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

2. Manual actions taken at the r.a.tO. cOntrol' .onsolConsole Center-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 Omanal reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the Console Center roactor control concolo 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 Console Center reactOr contro! coco!oc 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 trip. 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 Console Center reactor control conGelos (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 Console Centerro.acto control GOR60Gee6".

The plant response to the failure of an automatic eF manual-reactor trip 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 enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency Month 20XX TMI 3-64 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclea~r TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS I classification level will escalate to a Site Area Emergency via IC MSS_3. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent I the plant conditions needed to meet either IC MSS36 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.

IA reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA5

2. OP-TM-EOP-004, Lack of Primary-to-Secondary Heat Transfer

3. OP-TM-EOP-01 0, Emergency Procedure Rules, Guides And Graphs

4. OP-TM-EOP-001, Reactor Trip

5. 1102-4 Power Operation

6. OP-TM-641-000- Reactor Protection System (RPS/DSS)

Month 20XX TMI 3-65 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSU36 Initiating Condition:

Automatic or manual trip fails to shutdown the reactor.

Operating Mode Applicability:

1,2 Emergency Action Level (EAL):

I 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. Ar,-aAutomatic Ttrip did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

b. A-sSubsequent manual action taken at the rcactOr control concolocConsole Center is successful in shutting down the reactor.

OR

2. a. A-fAManual Ttrip did not shutdown the reactor as indicated by Reactor Power > 5%.

AND

b. EITHER of the followingt

-- J. ..... A----I ..... I---- ;--

-i- uDaseuentl manual aciion taken at -Rmurasaar cnuolm.- cn-nniec I

-S ! ;

SUEPE785 u " 6 t,

utt "iq 14 OWR Me FeaSION OR

2. A sSubsequent automatic Ttrip 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 trip that results in a reactor shutdown, and either a subsequent operator manual Month 20XX TMI 3-66 EP-AA-1009 (Revision XX)

Throm Mu.la li.nnri Qtafin Annoy F:valnn Nain-loar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS action taken at the Console Center reactor control consoles or an automatic trip 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 trip, operators will promptly initiate manual actions at the Console Center reactor control consolos to shutdown the reactor (e.g.,

initiate a manual reactor trip). 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.

EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly tako.n ua action at anotherl c-ation(s) on the reactor control consoles to'.roct. shutdown the (n._-._ initiato a man,-al macfor trio ,-*iflO a diffcrmnt c,;.tch\.. D n...ina.. .. un.........n. c.v.r.

Ifactors, tho.... initial or subsequent... eeffot f........... , ,.,....- ..-.-. .--.... --- *-*.......-- ÷-; '.c --.-.... concurren to manually trip the roactOr, or a concurrent plant plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manul- OF automatic trip 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 Console Center rcactOr control consolSG 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 trip). 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 Console Center reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip 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 Console Center reactor control consolcs are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC M.A36. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC MSA36 or FA1, an Unusual Event declaration is appropriate for this event.

I A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

I Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

Month 20XX TMI 3-67 EP-AA-1009 (Revision XX)

Throp Milo Inland qtafinn Annpv I=x*lnn Nu*IpAr M~ci ~I~d

~mtin An~'tExellnn Thrg~ Nuclea~r TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS

" 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 trip 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 trip signal and the trip 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.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU5

2. OP-TM-EOP-01 0, Emergency Procedure Rules, Guides And Graphs

3. OP-TM-EOP-001, Reactor Trip

4. 1102-4 Power Operation

5. OP-TM-641 -000- Reactor Protection System (RPS/DSS)

Month 20XX TMI 3-68 EP-AA-1009 (Revision XX)

Throm Uila lainnel Qtatinn Annoy F::alnn NiiIe-lar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSA42 Initiating Condition:

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

Operating Mode Applicability:

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

Note: The Emergency Director should declare the eventAled promptly upon determining that the applicable time 45-*i utes-has been exceeded, or will likely be exceeded.

1. a. AR,-UNPLANNED event results in the inability to monitor ANYone 9o Mor Table Mlf t4he followi.ng parameters from within the Control Room for >15 minutes O.lenge .

r[,,e t.able belo-w]

[ PA[R parameter list; Table M1 Control Room Parameters Reactor Power

" Reactor Power RGS Levei " PZR Level

" RCS Pressure RCS Pressure

In Core/Core Exit Temperature In Corc/Corc Exit T-empeatufe " Level in at least one OTSG Lev.'Ioin at es I (site specific " OTSG Emergency Feed Water Flow number) steam; ge-nerators Steam GeneratorAuxiliary or EneCem mooC e Fee A'A ,at F low.

AND

b. Any Table M2Wf the f*.w.i* transient eveRts in progress.

0 lft Is^rv%

iif r11%

^ ^ ft, r ^r trmrr"!% r ft-e-r IA .

.... r .....

A r- 4..4Ia l,.4 -n..~.4a. +k.. Or-O/ . .11 1 6A lA

! !,~

ot 'a" gro -- r .. . - - I - . UM W. "- "- . 70 Ct toct

.ReaeWF4p 0 RO:Tii;i "I" Month 20XX TMI 3-69 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Table M2 Significant Transients

" Automatic Turbine Runback >25% thermal reactor power

" Electrical Load Rejection >25% full electrical load

" Reactor Trip

" ESAS Actuation

" Thermal Power oscillations > 10%

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 ono Or moroany 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, 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-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.

Month 20XX TMI 3-70 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS I 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 I the ability to determine one o Fnor-anv of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In I addition, if all indication sources for one OF or-eany 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.

I 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 TMI 3-71 EP-AA-1 009 (Revision XX)

IExelon Nuclear Three Mile Islamnd S~tation Annex Exelon Nucleanr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSU42 Initiating Condition:

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

Operating Mode Applicability:

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

Note: The Emergency Director should declare the Unu-'-ual Evcntcvent promptly upon determining that the applicable time F-nii*uAes-has been exceeded, or will likely be exceeded.

a- A-UNPLANNED event results in the inability to monitor one OrFmoMr ANY Table M1 parameters from within the Control Room for > 15 minutes.

Table M1 Control Room Parameters

" Reactor Power

RCS Level

" RCS Pressure

" In Core/Core Exit Temperature

Level in at least one OTSG.

OTSG Emeraency Feed Water Flow 1.of the following Darametercs from within the Contro!1 Room for 15 minutoc OF4OAgef.

Month 20XX TMI 3-72 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month 20XX TMI 3-73 EP-AA-1009 (Revision XX)

I=xelon Nuclear Thr~a Milp 11--aland Station Annex Exelnn NucleIAar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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.

I As used in this EAL, an "inability to monitor" means that values for one Or morangy 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.

I 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.

I 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 one O mer-eany of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In I addition, if all indication sources for one Or morany 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.

I Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

I Escalation of the emergency classification level would be via IC MSA42.

Month 20XX TMI 3-74 EP-AA-1009 (Revision XX)

Three Mile Island Station Annex Exelon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Basis Reference(s):

1. NEI 99-01 Rev 6, SU2 Month 20XX TMI 3-75 EP-AA-1009 (Revision XX)

ExAIon Nuclear Thrpp Mile Island Stsation Annaex Exelon Nucle~ar TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSA59 Initiating Condition:

I Hazardous event affecting a SAFETY SYSTEM needd-required for the current operating mode.

Operating Mode Applicability:

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

( 1-.a

. The occurrence of ANY of the following hazardous events:

" Seismic event (earthquake)

" Internal or external flooding event

" High winds or tornado strike

" FIRE

EXPLOSION

- (,ito ,p ,,ific haza, dc)

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

a.4, Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed-required by Technical Specifications for the current operating mode.

OR b..2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure-neededrequired by Technical Specifications for the current operating mode.

Basis:

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if larqe quantities of smoke and heat are observed.

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high enerav lines or comoonents) or an electrical comoonent failure (caused bv short circuits.

grounding. arcing, etc.) should not automatically be considered an explosion. Such Month 20XX TMI 3-76 EP-AA-1009 (Revision XX)

Three Mile Islaind Stsatio~n Annex Exellon Nuclear TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS events may require a post-event inspection to determine if the attributes of an explosion are present.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGOS. These are typically systems classified as safety-related.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed-required for the current operating mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. Manual or automatic electrical isolation of safety eguipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.

EAL 4402.a Basis This EAL addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in se*yoe/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

EAL 1.b.2 2.b Basis This EAL addresses damage to a SAFETY SYSTEM component that is- required to be operable by Technical Specifications for the current operating mode, and is not in se wlee/operation or readily apparent through indications alone, of-as well as damage to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC FS1 or RASI.

If the EAL conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.

Basis Reference(s):

1. NEI 99-01, Rev 6 SA9 Month 20XX TMI 3-77 EP-AA-1009 (Revision XX)

Throm Uila lalanel Atatinn Annow l::p~lnn N~nlaar Thin. Mu. IQhanrI ~t~tinn Ann.v Fvp inn Niir~Iisnr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MSU64 Initiating Condition:

RCS leakage for 15 minutes or longer.

Operating Mode Applicability:

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

Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 1456 -intes-has been exceeded, or will likely be exceeded.

1. RCS unidentified or pressure boundary leakage gieateF-than

> 10 gpm for > 15 minutes. (cito spcificvalue) for 15 minutoc Or longer-.

OR

2. RCS identified leakage thaR->25 gpm for> 15.t..._...p..i*,,-p-f*-'... .uF

>eateF for 15.9Finutesor longer.

OR

3. Leakage from the RCS to a location outside containment >25 gpm for > 15 minutes.

groator than 25 gpmR for 15 minutes Or longer.

Basis:

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

This IC addresses RCS leakage which may be a precursor to a more significant event.

In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

EAL #1 and EAL #2 Basis These EALs are focused on a loss of mass from the RCS due to "unidentified leakage",

"pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).

EAL #3 Basis This EAL addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system.

These EALs thus apply to leakage into the containment, a secondary-side system (e.g.,

steam generator tube leakage) or a location outside of containment.

Month 20XX TMI 3-78 EP-AA-1009 (Revision XX)

ExAIon Nuclear Threea Mile Island Station Annex Exelon Nuclanr TABLE TMI 3-2: EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via lCs of Recognition Category RA or F.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU4

2. OP-TM-220-251 RCS Leak Rate Determination

3. OP-TM-220-252, Primary - To - Secondary Leakrate Determination

4. OP-TM-PRF1-0405 RB Sump Level HI

5. UFSAR 6.4.3, Bases of Leakage Estimate

6. UFSAR 6.4.4, Design Basis Leakage

7. OP-TM-AOP-050, Reactor Coolant Leakage

8. Technical Specification 3.1.6, Leakage and Table 4.1-2, Minimum Equipment Test Frequency Month 20XX TMI 3-79 EP-AA-1009 (Revision XX)

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