Text

Attachment 8 - EP-AA-1010, Revised Radiological Emergency Plan Annex Information for Oyster Creek Nuclear Generating Station
	ML15072A399
Person / Time
Site:	Oyster Creek
Issue date:	03/02/2015
From:	; Exelon Generation Co
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML15071A122	List: ML15071A119; ML15071A128; ML15071A137; ML15071A172; ML15071A173; ML15072A388; ML15072A389; ML15072A399; ML15072A411; RA-15-019, Attachment 1 - Response to the Request for Additional Information - 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-15-019, Attachment 11 - EP-AA-1009, Revised Radiological Emergency Plan Annex Information for Three Mile Island Station; RA-15-019, Attachment 3 - EP-AA-1002, Revised Radiological Emergency Plan Annex Information for Byron Station; RA-15-019, Attachment 4 - EP-AA-1003, Revised Radiological Emergency Plan Annex Information for Clinton Power Station; RA-15-019, Attachment 5 - EP-AA-1004, Revised Radiological Emergency Plan Annex Information for Dresden Station; RA-15-019, Attachment 6 - EP-AA-1005, Revised Radiological Emergency Plan Annex Information for LaSalle County Station; RA-15-019, Attachment 7 - EP-AA-1008, Revised Radiological Emergency Plan Annex Information for Limerick Generating Station; RA-15-019, Attachment 9 - EP-AA-1007, Revised Radiological Emergency Plan Annex Information for Peach Bottom Atomic Power Station; RA-15-019, Exelon - Response to Request for Additional Information 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-15-082, Response to Request for Additional Information 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; TMI-15-034, Attachment 8 - EP-AA-1010, Revised Radiological Emergency Plan Annex Information for Oyster Creek Nuclear Generating Station;
References
	RA-15-019, RS-15-082, TMI-15-034 EP-AA-1010
	Download: ML15072A399 (325)
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ATTACHMENT 8 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR OYSTER CREEK NUCLEAR GENERATING STATION EP-AA-1 010 Enclosures S 0 0 Enclosure 8A -Revised EAL Comparison Matrix Document Enclosure 8B -Revised EAL Red-Line Basis Documents Enclosure 8C -Revised EAL Basis Documents NEI 99-01 Rev 6 1 Proposed EAL Justification AGI Initiating Condition

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

All Example Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the General Emergency 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.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

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

RGI Initiating Condition:

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

1,2,3,4,D Emergency Action Level (EAL): Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.C Olassification 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. 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. >1000 mRemTEDE OR b. > 5000 mRem CDE Thyroid OR 3. Field survey results at or beyond the site boundary indicate EITHER: a. Gamma (closed window) dose rates >1000 mR/hr are expected to continue for > 60 minutes.OR b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

Table RI Effluent Monitor Thresholds Effluent Monitor General Emergency Main Stack RAGEMS 8.69 E+01 uCi/cc HRM OR 3.53 E-08 amps HRM Turbine Building RAGEMS 7.17 E-01 uCi/cc HRM HRM = High Range Monitor D1 No Change E Difference 1: Deviation Listed site-specific monitors and Threshold values to ensure timely classification.

1) Added the following to bullet #3 " Classification based on effluent monitor readings assumes that a release path to the environment is established." In order to delete the following from the basis "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." This allows for more timely classification since all the basis information pertaining to Note bullet 3 will be contained in the IC and therefor readily available on the 1 x17 procedure matrix used by the SM.Page 1 of'67 NEI 99-01 Rev 6 1 Proposed EAL ] Justification ASI Initiating Condition

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

All Example Emergency Action Levels: (1 or 2 or 3)Notes:* The Emergency Director should declare the Site Area Emergency 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.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

1. .Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer: (site specific monitor list and threshold values)2. Dose assessment actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose receptor point)3. Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point): " Closed window dose rates greater than 100 mRlhr expected to continue for 60 minutes or longer.* Analysis of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

RS1 Initiating Condition:

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

1,2,3,4,D Emergency Action Level (EAL): Notes: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* Classification based on effluent monitor readings assumes that a release 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. 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. >100 mRem TEDE OR b. > 500 mRem CDE Thyroid OR 3. Field survey results at or beyond the site boundary indicate EITHER: a. Gamma (closed window) dose rates >100 mR/hr are expected to continue for > 60 minutes.OR b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.

Table Ri Effluent Monitor Thresholds Effluent Monitor Site Area Emergency Main Stack RAGEMS 8.69 E+00 uCi/cc HRM OR 3.53 E-09 amps HRM Turbine Building RAGEMS 3.65 E+05 cpm LRM HRM = High Range Monitor LRM = Low Range Monitor D No Change E Difference 1 Deviation Listed site-specific monitors and Threshold values to ensure timely classification.

1) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order to delete the following from the basis "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." This allows for more timely classification since all the basis information pertaining to Note bullet 3 will be contained in the IC and therefor readily available on the 11 x17 procedure matrix used by the SM.Page 2 of 67 NEI 99-01 Rev 6 ] Proposed EAL I Justification AAI Initiating Condition

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

All Example Emergency Action Levels: (1 or 2 or 3)Note:* The Emergency Director should declare the Alert 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.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer: (site-specific monitor list and threshold values)2. Dose assessment actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose receptor point)3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point) for one hour of exposure.4. Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point): " Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.* Analysis of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

RAI Initiating Condition:

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

1,2,3,4, D Emergency Action Level (EAL): Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.C Olassification 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. 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 a 60 minutes.OR b. Analyses of field survey sam ples in d ic ate > 50 mRem CDE Thyroid for 60 minutes of inhalation.

Table R1 Effluent Monitor Thresholds Effluent Monitor Alert Main Stack RAGEMS 8.69 E-01 uCi/cc HRM OR 3.53 E-10 amps HRM Turbine Building RAGEMS 3.65 E+04 cpm LRM HRM = High Range Monitor LRM = Low Range Monitor D1 No Change E Difference 1 Deviation Listed site-specific monitors and Threshold values to ensure timely classification.

1) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order to delete the following from the basis "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." This allows for more timely classification since all the basis information pertaining to Note bullet 3 will be contained in the IC and therefor readily available on the 1 lx17 procedure matrix used by the SM.2) Calculation was performed to determine the radiation monitor response for a radioactive liquid release with an activity equivalent to provide 1 Omrem TEDE or 50mrem thyroid CDE at the site boundary via the normal site release pathway. The calculation determined the liquid radwaste system was no longer used and was cut and capped oiff. As such the liquid radwaste effluent monitor was not included in this EAL Page 3 of 67 NEI 99-01 Rev 6 [ Proposed EAL I Justification AU1 Initiating Condition

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

All Example Emergency Action Levels: (1 or 2 or 3 Note:* The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes 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.* If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, 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: (site-specific monitor list and threshold values corresponding to 2 times the controlling document limits)2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.3. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times (site-specific effluent release controlling document limits) for 60 minutes or longer.RUI Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability:

1,2,3,4, D Emergency Action Level (EAL): Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 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. Readings on ANY Table R1 Effluent Monitor > Table R1 value for>60 minutes: Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Event Main Stack RAGEMS 4.07 E+03 cps LRM Turbine Building RAGEMS 4.16 E+02 cpm LRM LRM = Low Range Monitor OR 2. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.D-1 No Change E Difference 1 Deviation Listed site-specific monitors and Threshold values to ensure timely classification.

1) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order to delete the following from the basis "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." This allows for more timely classification since all the basis information pertaining to Note bullet 3 will be contained in the IC and therefor readily available on the 1 1x17 procedure matrix used by the SM.2) Calculation was performed to determine the radiation monitor response for a radioactive liquid release with an activity equivalent to provide 2 times the ODCM.The calculation determined the liquid radwaste system was no longer used and was cut and capped oiff. As such the liquid radwaste effluent monitor was not included in this EAL Page 4 of67 NEI 99-01 Rev 6 Proposed EAL Justification AU2 Initiating Condition

-- GENERAL EMERGENCY Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer Operating Mode Applicability:

All Example Emergency Action Levels: NOTES: The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded.or will likely be exceeded 1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer.RG2 Initiating Condition:

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

1,2,3,4, D Emergency Action Level (EAL): Note: The Emergency Director should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for 60 minutes or longer.D No Change W-Difference 1 1 Deviation 1) EAL will be used in accordance with the discussion in Section 1.4, NRC Order EA-12-051, it is recommended that this EAL be implemented when the enhanced spent fuel pool level instrumentation is available for use. The completion of the 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 Report to March 12, 2012 Commission Order Modifying Licenses with Regard for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated October 25,2012.Page 5 of 67 NEI 99-01 Rev 6 Proposed EAL Justification AS2 RS2 NChne- Dfeec Deato Initiating Condition

-SITE AREA EMERGENCY Initiating Condition:

RS1 No Change Difference Deviation Spent fuel pool level at (site-specific Level 3 description)

Spent fuel pool level at (site-specific Level 3 description).

1 ) EAL will be used in accordance with the discussion in Section 1.4, NRC Operating Mode Applicability:

All Operating Mode Applicability:

Order EA-12-051, it is recommended that this EAL be implemented when the 1, 2, 3, 4, D enhanced spent fuel pool level instrumentation is available for use. The Example Emergency Action Levels: completion of the enhanced SFP level indicators and need for the inclusion of this Emergency Action Level (EAL): EAL is being tracked in accordance with Exelon Generation Company, LLC's 1. Lowering of spent fuel pool level to (site-specific Level 3 value). Lowering of spent fuel pool level to (site-specific Level 3 value). Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated October 25,2012.Page 6 of 67 NEI 99-01 Rev 6 Proposed EAL Justification AA2 Initiating Condition

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

All Example Emergency Action Levels: (1 or 2 or 3 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors: (site-specific listing of radiation monitors, and the associated readings, setpoints and/or alarms)3. Lowering of spent fuel pool level to (site-specific Level 2 value).Initiating Condition:

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

1,2,3,4, D Emergency Action Level (EAL): 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 mRem/hr.OR 3. Lowering of spent fuel pool level to (site-specific Level 2 value).D No Change E Difference 1 1 Deviation 1) Listed site-specific monitors and Threshold values to ensure timely classification.

2) EAL #3 will be used in accordance with the discussion in Section 1.4, NRC Order EA-12-051, it is recommended that this EAL be implemented when the enhanced spent fuel pool level instrumentation is available for use. The completion of the 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 Report to March 12, 2012 Commission Order Modifying Licenses with Regard for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated October 25,2012.Table R2 Refuel Floor ARM's* C-5, Crit Mon* C-10, North Wall* C-9, North Wall* B-9, Open Floor Page 7 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification AU2 Initiating Condition:

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

All Example Emergency Action Levels: 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following: (site-specific level indications).

AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.(site-specific list of area radiation monitors)RU2 Initiating Condition:

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

1,2,3,4, D Emergency Action Level (EAL): 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

Refueling Cavity water level < 583 inches (GEMAC Wide Range, floodup calibration).

OR* Indication or report of a drop in water level in the REFUELING PATHWAY.AND b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R2.No Change E Difference 1:1 Deviation Listed site specific level indication and monitors to ensure timely classification.

Table R2 Refuel Floor ARM's* C-5, Crit Mon* C-10, North Wall* C-9, North Wall* B-9, Open Floor Page 8 of 67 NEI 99-01 Rev 6 1 Proposed EAL [ Justification AA3 Initiating Condition

-ALERT Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.Operating Mode Applicability:

All Example Emergency Action Levels: (1 or 2)Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted 1. Dose rate greater than 15 mRlhr in ANY of the following areas:* Control Room* Central Alarm Station* (other site-specific areas/rooms)

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

RA3 Initiating Condition:

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

1,2, 3, 4,D Emergency Action Level (EAL): Note: If the equipment in the room or area listed in Table R3 was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted.

1. Dose rate > 15 mR/hr in ANY of the following Table R3 areas: Table R3 Areas Requiring Continuous Occupancy" Main Control Room" Central Alarm Station -(by survey)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 51' elevation Modes 3 and 4 No Change 1-- Difference 1 Deviation Listed site specific plant rooms and areas with identified mode applicability to ensure timely classification.

Page 9 of 67 NEI 99-01 Rev 6 Proposed EAL [ Justification SU3 Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: 1. (Site-specific radiation monitor) reading greater than (site-specific value).OR 2. Sample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.

RU3 Initiating Condition:

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

1,2 Emergency Action Level (EAL): 1. Offgas system radiation monitor Hi-Hi alarm.OR 2. Specific coolant activity > 4.0 uClIgm Dose equivalent 1-131.H No Change E Difference 1 Deviation 1 ) Listed site-specific monitor and Threshold value to ensure timely classification.

2) Listed this system category EAL in the radiological category EAL section to 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 radiological category vice system category.Page 10 of 67 NEt 99-01 Rev 6 Fission Product Barrier Matrix FG1 Loss of any two barriers AND Loss or Potential Loss of third barrier.1,2 FS1 Loss or Potential Loss of ANY two barriers.1,2 FAl ANY Loss or ANY Potential Loss of either Fuel Clad or RCS 1,2 FC -Fuel Clad RC -Reactor Coolant System CT -Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss 1. RCS Activity A. UNPLANNED rapid drop in primary A. Primary containment pressure greater.rivity containment pressure following than (site-specific value)Continment A. (Site specific indications that Containment cl acivit i s grat A. Primary containment pressure pmmary containment pressure rise OR Pressure/

reactor coolant activity is greater None greater than (site-specific value) None OR B. (site-specific explosive mixture) exists Pressre than 300 uCi/gm dose equivalent due to RCS leakage. B. Primary containment pressure inside primary containment Primary 111 .Piaycnanetpesr Containment

31) response not consistent with LOCA OR Conditions conditions.

C. HCTL exceeded.A. RPV water level cannot be restored A. RPV water level cannot be A. Pnmary containment flooding and maintained above (site-specific restored and maintained above 2. RPV Water required.

RPV water level corresponding to (site-specific RPV level None None A. Primary containment flooding required.Level top of active fuel) or cannot be corresponding to the top of active determined.

fuel) or cannot be determined.

A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR A. UNISOLABLE primary system leakage A. UNISOLABLE break in any of the that results in exceeding EITHER of B. Intentional primary containment A.C UNeakBL bRkantnyofth the foloing: venting per FOPs 3,RCS Leak Rate/ following: ( site-specific systems with 1. Mollowmng:

venting OR Primary None None potential for high-energy ine Max Normal Operating ORNone Containment OR Temperature C. UNISOLABLE primary system Isolation Failure OR leakage that results in exceeding eB. Emergency RPV Depressurzation

2. Max Normal Operating Area EITHER of the following:

Radiation Level. 1. Max Safe Operating Temperature.

OR 2. Max Safe Operating Area Radiation Level.4.Primary A. Primary Containment Radiation A. Primary Containment Radiation A. Primary Containment Radiation Monitor Containment Monitor reading greater than None Monitor reading greater than (site- None None reading greater than (site-specific Radiation (site-specific value), specific value). value).A. Any Condition in the opinion of 5. Emergency the Emergency Director that A. Any Condition in the opinion of the A. ANY Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the Director Judgment indicates Loss of the Fuel Clad Emergency Director that indicates Emergency Director that indicates Loss Emergency Director that indicates Emergency Director that indicates Loss Emergency Director that indicates Potential DirectorJudgment inBarrier.

Potential Loss of the Fuel Clad Barrier. of the RCS Barrier. Potential Loss of the RCS Barrier. of the Containment Barrier. Loss of the Containment Barrier.Page I I of 67 Proposed Fission Product Barrier Matrix Fission Product Barrier Matrix Hot Matrix GENERAL EMERGENCY SITE AREA EMERGENCY ALERT FG1 Loss of any two barriers AND Loss or Potential Loss of third barrier.m0 FSi Loss or Potential Loss of ANY two barriers.E[A FAl ANY Loss or ANY Potential Loss of either Fuel Clad or RCS m0 FC -Fuel Clad RC -Reactor Coolant System CT -Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss 1. RCS Activity Coolant activity *300 uCil/gm Dose None None Nose None None Equivalent 1-131 2. RPV water level cannot be restored and 1. RPV water level cannot be restored and 2. RPV Water 1. Plant conditions indicate prmary maintained Inches TA maintained Inches Plant conditions indicate primary containment Level containment flsoding is required.

OR OR None None flooding is required.3. RPV water level cannot be determined.

2. RPV water level cannot be determined.

3. Drywell pressure > 44 pslg and rising.OR 1. Drwell pressure >3.0 psig. 1. UNPLANNED rapid drop in Drywell pressure 4. a. Drywell or tows hydrogen concentration

3. Primary AND following Drywell pressure rise. > 6%.3.Pimr ADAND Containment None None 2. Drywell pressure rise is due to RCS None 2. ORD Pressure/Conditions leakage 2. Drywell pressure response not consistent

b. Drywell or torus oxygen concentration with LOCA conditions.

> 5%.OR 5. Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded.3. UNISOLABLE primary system leakage that 1. UNISOLABLE Main Steam Line (MSL), results in EITHER of the following:

Isolation Condenser, Feedwater, or a. Secondary Containment area RWCU line break, temperature

> EMG-3200.11 Max 4.RCS Leak Rate None None OR Normal (Table 11) operating level. None None OR 2. Emergency RPV Depressurization is b. Secondary Containment area radiation required, level > EMG-3200.11 Max Normal (Table 12) operating level.5.Primary Containment Hi Range Radiation Containment Hi Range Radiation Monitoring Containment Monitoring System (CHRRMS) reading None System (CHRRMS) reading > 160R/hr. None None Containment Hi Range Radiation Monitoring Radiation

> 530 R/hr. SySystem (CHRRMS) reading > 1210 R/hr.1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.OR 2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

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

a. Secondary Containment area temperature s EMG-3200.11 Max Safe (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11 Max Safe (Table 12) operating level.7. Emergency A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the A. Any Condition in the opinion of the Emergen Emergency Director that indicates Loss Emergency Director that indicates Potential Emergency Director that indicates Loss of Emergency Director that indicates Potential Emergency Director that indicates Loss of the Emergency Director that indicates Potential Loss Director Judgment of the Fuel Clad Barrier. Loss of the Fuel Clad Barrier. the RCS Barrier. Loss of the RCS Barrier. Containment Barrier. of the Containment Barrier.Page 12 of67 NEI 99-01 Rev 6 FC1 Category:

Fuel Clad Barrier RCS Activity Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. (Site specific indications that reactor coolant activity is greater than 300 uCi/gm dose equivalent 1-131)Proposed EAL Justification

-1 +FCl Category:

Fuel Clad Barrier RCS Activity Operating Mode Applicability:

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

Loss Coolant activity > 300 uCi/gm Dose Equivalent 1-131.E No Change 1 Difference 1 Deviation 1) Listed site-specific threshold value to ensure timely classification.

Page 13 of167 NEI 99-01 Rev 6 [ Proposed EAL Justification FC2 Category:

Fuel Clad Barrier RPV Water Level Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Primary containment flooding required.Potential Loss A. RPV water level cannot be restored and maintained above (site-specific RPV water level corresponding to top of active fuel) or cannot be determined.

FC2 Category:

Fuel Clad Barrier RCS Activity Operating Mode Applicability:

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

Loss 1. Plant conditions indicate primary containment flooding is required.Potential Loss 2. RPV water level cannot be restored and maintained

> 0 inches TAF OR 3. RPV water level cannot be determined.

E No Change 1 Difference

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

Page 14 of 67 NEI 99-01 Rev 6 1 Proposed EAL I Justification FC4 Category:

Fuel Clad Barrier Primary Containment Radiation Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Primary Containment Radiation Monitor reading greater than (site-specific value).FC5 Category:

Fuel Clad Barrier Primary Containment Radiation Operating Mode Applicability:

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

Loss Containment Hi Range Radiation Monitoring System (CHRRMS) reading> 530 R/hr.E No Change 1 Difference 1 Deviation 1 ) Listed site-specific monitor and threshold value to ensure timely classification.

Page 15 of 67 NEI 99-41 Rev 6 ] Proposed EAL Justification FC6 Category:

Fuel Clad Barrier Emergency Director Judgment Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.FC7 Category:

Fuel Clad Barrier Emergency Director Judgment Operating Mode Applicability:

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

Loss 1. Any Condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.Potential Loss 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.E No Change El Difference 1-" Deviation Page 16 of67 NEI 99-01 Rev 6 Proposed EAL Justification RC1 Category:

Reactor Coolant System Barrier Primary Containment Pressure Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Primary containment pressure greater than (site-specific value) due to RCS leakage.RC3 Category:

Reactor Coolant System Barrier Primary Containment Pressure/Conditions Operating Mode Applicability:

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

Loss 1. Drywell pressure > 3.0 psig.AND 2. Drywell pressure rise is due to RCS leakage E No Change 1 Difference 1: Deviation 1) Listed site-specific threshold value to ensure timely classification.

Page 17 of67 NEI 99-01 Rev 6 [ Proposed EAL Justification RC2 Category:

Reactor Coolant System Barrier RPV Water Level Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

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

RC2 Category:

Reactor Coolant System Barrier RPV Water Level Operating Mode Applicability:

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

Loss 1. RPV water level cannot be restored and maintained

> 0 Inches TAF OR 2. RPV water level cannot be determined.

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

Page 18 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification RC3 Category:

Reactor Coolant System Barrier RCS Leak Rate Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. UNISOLABLE break in any of the following: ( site-specific systems with potential for high-energy line breas)OR B. Emergency RPV Depressurization Potential Loss A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Normal Operating Temperature OR 2. Max Normal Operating Area Radiation Level.RC4 Category:

Reactor Coolant System Barrier RCS Leak Rate Operating Mode Applicability:

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

Loss 1. UNISOLABLE Main Steam Line (MSL), Isolation Condenser, Feedwater, or RWCU line break.OR 2. Emergency RPV Depressurization is required.Potential Loss 3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature

> EMG-3200.11 Max Normal (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11 Max Normal (Table 12) operating level.W No Change 1 Difference 1 Deviation 1) Listed site-specific systems and threshold values to ensure timely classification.

Page 19 of 67 NEI 99-01 Rev 6 1 Proposed EAL I Justification RC4 Category:

Reactor Coolant System Barrier Primary Containment Radiation Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Primary Containment Radiation Monitor reading greater than (site-specific value).RC5 Category:

Reactor Coolant System Barrier Primary Containment Radiation Operating Mode Applicability:

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

Loss Containment Hi Range Radiation Monitoring System (CHRRMS) reading> 1000 R/hr.-No Change [: Difference 1 Deviation 1) Listed site-specific monitor and threshold value to ensure timely classification.

Page 20 of 67 NEI 99-01 Rev 6 Proposed EAL ] Justification RC6 Category:

Reactor Coolant System Barrier Emergency Director Judgment Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.RC7 Category:

Reactor Coolant System Barrier Emergency director Judgment Operating Mode Applicability:

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

Loss 1. ANY Condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.Potential Loss 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.No Change I: Difference 1 Deviation Page 21 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification CT1 Category:

Containment Barrier Primary Containment Conditions Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss C. UNPLANNED rapid drop in primary containment pressure following primary containment pressure rise OR B. Primary containment pressure response not consistent with LOCA conditions.

Potential Loss D. Primary containment pressure greater than (site-specific value)OR E. (site-specific explosive mixture) exists inside primary containment OR 3. HCTL exceeded.CT3 Category:

Containment Barrier Primary Containment Pressure/Conditions Operating Mode Applicability:

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

Loss 1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.OR 2. Drywell pressure response not consistent with LOCA conditions.

Potential Loss 3. Drywell pressure > 44 psig and rising.OR 4. a. Drywell or torus hydrogen concentration

> 6%.AND b. Drywell or torus oxygen concentration

> 5%.OR 5. Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded.D No Change Difference Deviation 1) Listed site-specific threshold values to ensure timely classification.

2) The words "and rising" were added to account for the momentary spike in pressure where pressure is now lowering, the risk of a potential loss of containment is no longer present, this wording is also consistent with present EAL wording.Page 22 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification CT2 Category:

Containment Barrier RPV Water Level Operating Mode Applicability:

Power Operation.

Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Potential Loss A. Primary containment flooding required.Category:

Containment Barrier RPV Water Level Operating Mode Applicability:

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

Potential Loss Plant conditions indicate primary containment flooding is required.CT2 LI No Change Difference D Deviation Page 23 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification CT3 Category:

Containment Barrier Primary Containment Isolation Failure Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR B. Intentional primary containment venting per EOPs OR C. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR 2. Max Safe Operating Area Radiation Level.CT6 Category:

Containment Barrier Primary Containment Isolation Failure Operating Mode Applicability:

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

Loss 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.OR 2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

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

a. Secondary Containment area temperature

> EMG-3200.11 Max Safe (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11I Max Safe (Table 12) operating level.W No Change EI Difference 1 Deviation 1) Listed site-specific threshold values to ensure timely classification.

Page 24 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CT4 Category:

Containment Barrier Primary Containment Radiation Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Potential Loss A. Primary Containment Radiation Monitor reading greater than (site-specific value).CT5 Category:

Containment Barrier Primary Containment Radiation Operating Mode Applicability:

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

Potential Loss Containment Hi Range Radiation Monitoring System (CHRRMS) reading> 1210 R/hr.7- No Change 1 Difference 1 Deviation 1 ) Listed site-specific monitor and threshold value to ensure timely classification.

Page 25 of 67 NEI 99-01 Rev 6 Proposed EAL I Justification CT6 Category:

Containment Barrier Emergency director Judgment Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Fission Product Barrier Threshold:

Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.CT7 Category:

Containment Barrier Emergency Director Judgment Operating Mode Applicability:

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

Loss 1. Any Condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.Potential Loss 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.W No Change E Difference E Deviation Page 26 of 67 NEI 99-01 Rev 6 Proposed EAL Justification SG1I Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that (site-specific hours) has been exceeded, or will likely be exceeded.1. a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).AND b. EITHER of the following: " Restoration of at least one emergency bus in less than (site-specific hours) is not likely." (Site-specific indication of an inability to adequately remove heat from the core)MG1 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1 Loss of ALL offsite AC power to 4160V Buses 1C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and ID.AND 3. EITHER of the following:

a. Restoration of at least one 4160V Bus (1C or 1 D) in < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is not likely.OR b. RPV water level cannot be restored and maintained

> -20 inches TAF.I -1 No Change I Difference

[IDeviation Listed site specific equipment, site specific time based on station blackout coping analysis, and site specific indication to ensure timely classification.

Page 27 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification SS1 Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses)for 15 minutes or longer.MS1 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of ALL offsite AC Power to 4160V Buses 1 C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D.AND 3. Failure to restore power to at least one 4160V Bus (1 C or 1D) in < 15 minutes from the time of loss of both offsite and onsite AC power.F-1 No Change I Difference FI Deviation 1) Listed site specific equipment to ensure timely classification.

Page 28 of 67 NEI 99-01 Rev 6 Proposed EAL Justification SA1 Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.1. a. AC power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer.AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.MA1 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note:-The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. AC power capability to 4160V Buses 1 C and 1D reduced to only one of the following power sources for > 15 minutes.* Startup Transformer SA* Startup Transformer SB* EDG-1 Emergency Diesel Generator* EDG-2 Emergency Diesel Generator AND 2. ANY additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.No Change i Difference F Deviation 1) Listed site specific equipment to ensure timely classification.

_____________________________________________________

L Page 29 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification Sul Initiating Condition:

UNUSUAL EVENT Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes 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 MU1 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Loss of ALL offsite AC power capability to 4160V Buses 1C and 1D for_>15 minutes.Fx] No Change F Difference F- Deviation 1) Listed site specific equipment to ensure timely classification.

Page 30 of 67 NEI 99-01 Rev 6 Proposed EAL Justification bud Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes 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.AND Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific vital DC buses) for 15 minutes or longer.MG2 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of ALL offsite AC power to 4160V Busesl C and 1 D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D..AND 3. Voltage is < 115 VDC on 125 VDC battery busses B and C.AND 4. ALL AC and Vital DC power sources have been lost for_> 15 minutes.D] No Change F Difference F Deviation 1) Listed site specific equipment to ensure timely classification.

2) Removed the word "indicated" this will allow for an indication problem to not cause confusion on the need to declare.Page 31 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.Indicated voltage is less than (site-specific bus voltage value) on ALL Vital DC buses for 15 minutes or longer.MS2 Initiating Condition:

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

D No Change M Difference F Deviation 1) Listed site specific equipment and site specific value to ensure timely classification.

2) Removed the word "indicated" this will allow for an indication problem to not cause confusion on the need to declare.1,2,3 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Voltage is < 115 VDC on 125 VDC battery busses B and C for> 15 minutes.Page 32 of 67 NEI 99-01 Rev 6 Proposed EAL Justification SS5 Initiating Condition:

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

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

AND c. EITHER of the following conditions exist: 1. (Site-specific indication of an inability to adequately remove heat from the core)OR 2. (Site-specific indication of an inability to adequately remove heat from the RCS)MS3 Initiating Condition:

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

1,2 Emergency Action Level (EAL): 1. Automatic scram did not shutdown the reactor as indicated by Reactor Power> 2%.AND 2. ALL manual I ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 2%.AND 3. EITHER of the following conditions exist: " RPV water level cannot be restored and maintained

> -20 inches TAF.OR" Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded.D No Change I Difference FIDeviation

1) Listed site specific indications to ensure timely classification.

2) Mode 2 included in operating mode applicability as per developer notes.3) Added ARI as an equivalent scram Page 33 of 67 NEI 99-01 Rev 6 Proposed EAL Justification SA5 Initiating Condition:

ALERT Automatic or manual (trip [PWR] / scram [BWR]) 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:

Power Operation Example Emergency Action Levels: 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. An automatic (trip [PWR] / scram [BWR]) did not shutdown the reactor.AND b. Manual action taken at the reactor control consoles are not successful in shutting down the reactor.MA3 Initiating Condition:

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

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

1. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 2%.D No Change D Difference

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

2) Mode 2 included in operating mode applicability as per developer notes.3) Added ARI as an equivalent scram Page 34 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification SU5 Initiating Condition:

UNUSUAL EVENT Automatic or manual (trip [PWRJ / scram [BWRJ) fails to shutdown the reactor.Operating Mode Applicability:

Power Operation Example Emergency Action Levels: (1 or 2)Note: A manual action is any operator action, or set of actions, which causes 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. An automatic (trip [PWRJ / scram [BWRJ) did not shutdown the reactor.AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.2. a. A manual scram ([PWRJ / scram (BWR]) did not shutdown the reactor.AND 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 [PWRJ / scram [BWR]) is successful in shutting down the reactor.MU3 Initiating Condition:

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

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

1. a. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND b. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.OR 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND b. EITHER of the following:

1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.OR 2. Subsequent automatic scram / ARI is successful in shutting down the reactor.D No Change M Difference F 1Deviation

1) Listed site specific indications to ensure timely classification.

2) Mode 2 included in operating mode applicability as per developer notes.3) Added ARI as an equivalent scram Page 35 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification SA2 Initiating Condition:

ALERT UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.1. a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.[see table below][ BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In Core/Core Exit Temperature nPool Level Levels in at least (site specific Suppression Pnumber) steam generators Steam Generator Auxiliary or Suppression Pool Temperature Emergency Feed Water Flow AND b. Any of the following transient events in progress.* Automatic or Manual runback greater than 25% thermal reactor power* Electrical load rejection greater than 25% full electrical load* Reactor Scram [BWRJ / trip [PWRJ* ECCS (SI) actuation* Thermal power oscillations greater than 10% [BWR]MA4 Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. UNPLANNED event results in the inability to monitor ANY Table M1 parameters from within the Control Room for > 15 minutes.W No Change FiDifference F Deviation Table M1 Control Room Parameters

Reactor Power* RPV Water Level* RPV Pressure* Drywall Pressure* Torus Water Level* Torus Water Temperature AND 1 2. ANY Table M2 transient in progress.Table M2 Significant Transients

Turbine Trip* Reactor Scram* ECCS Actuation* Thermal power change > 25%.Thermal Power oscillations

> 10%Page 36 of 67 NEI 99-01 Rev 6 Proposed EAL I Justification SU2 Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.[see table below][ BWR parameter list] [PWR parameter list]Reactor Power Reactor Power RPV Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In Core/Core Exit Temperature nPool Level Levels in at least (site specific Suppression Pnumber) steam generators nPool Temperature Steam Generator Auxiliary or Suppression PEmergency Feed Water Flow MU4 Initiating Condition:

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

1,2, 3 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.UNPLANNED event results in the inability to monitor ANY Table M1 parameters from within the Control Room for > 15 minutes.I No Change EIDifference F- Deviation Table M1 Control Room Parameters

Reactor Power* RPV Water Level* RPV Pressure* Drywell Pressure* Torus Water Level* Torus Water Temperature Page 37 of67 NEI 99-01 Rev 6 1 Proposed EAL Justification SA9 Initiating Condition:

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

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

Internal or external flooding event* High winds or tornado strike* FIRE* EXPLOSION* (site-specific hazards)* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.MA5 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note: If it is determined that the conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.1. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)" Internal or external flooding event" High winds or tornado strike" FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager AND 2. EITHER of the following:

a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.OR b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical Specifications for the current operating mode.FH No Change I Difference F Deviation 1) No additional site specific hazards noted 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 minimize confusion.

3) Added additional note to easily direct the operator to lesser ICs Page 38 of 67 NEI 99-01 Rev 6 [ Proposed EAL Justification SU4 Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes 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.2. RCS identified leakage greater than (site-specific value) for 15 minutes or longer 3. Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer MU6 Initiating Condition:

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

1,2,3 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for >15 minutes.OR 2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.OR 3. Leakage from the RCS to a location outside the Drywell >25 gpm for> 15 minutes.FD 1 No Change -IDifference F-1 Deviation 1) Listed site specific values to ensure timely classification.

2) in EAL #3 Changed wording from containment to Drywell for clarity to better define the primary containment structure.

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

Page 39 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification Initiating Condition:

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

Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: (1 or 2 or 3)1. Loss of ALL of the following onsite communication methods: (site-specific list of communications methods)2. Loss of ALL of the following ORO communication methods: (site-specific list of communications methods)3. Loss of ALL of the following NRC communication methods: (site-specific list of communications methods)SU6 MU7 Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

1,2,3 Emergency Action Level (EAL): No Change EIDifference F Deviation 1) Listed site specific communication methods to ensure timely classification.

1. Loss of all Table M3 Onsite communications capability affecting the ability to perform routine operations.

OR 2. Loss of all Table M3 Offsite communication capability affecting the ability to perform offsite notifications.

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

Table M3 Communications Capabil ty System Onsite Offsite NRC Plant Paging System X Station Radio X Conventional Telephone lines X X X Satellite Phone system X X NARS X HPN X X ENS X X Page 40 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CA2 Initiating Condition:

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

Cold Shutdown, Refueling, Defueled Example Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses)for 15 minutes or longer.CA1 Initiating Condition:

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

3,4, D Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of all offsite AC power to 4160V Buses 1C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D..AND 3. Failure to restore power to at least one 4160V bus (1C or 1D) in < 15 minutes from the time of loss of both offsite and onsite AC power.E No Change I Difference 1:1 Deviation 1) Listed site specific equipment to ensure timely classification.

Page 41 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CU2 Initiating Condition:

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

Cold Shutdown, Refueling, Defueled Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.1. a. AC power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer.AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.Cul Initiating Condition:

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

3,4, D Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. AC power capability to 4160V Buses 1C and 1 D reduced to only one of the following power sources for > 15 minutes." Startup Transformer SA* Startup Transformer SB" EDG-1 Emergency Diesel Generator" EDG-2 Emergency Diesel Generator AND 2. ANY additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.E No Change 1 Difference 1 Deviation 1) Listed site specific equipment to ensure timely classification.

Page 42 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CA6 Initiating Condition

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

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

Internal or external flooding event* High winds or tornado strike" FIRE* EXPLOSION* (site-specific hazards)* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.CA2 Initiating Condition:

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

3.4 Emergency

Action Level (EAL): Note: If it is determined that the conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.1 .The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

Internal or external flooding event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager AND 2. EITHER of the following:

a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.OR b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical Specifications for the current operating mode.-1D No Change E Difference Deviation 1) No additional site specific hazards noted 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 minimize confusion.

3) Added additional note to easily direct the operator to lesser Cs Page 43 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CU4 Initiating Condition:

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

Cold Shutdown, Refueling Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.Indicated voltage is less than (site-specific bus voltage value) on required Vital DC buses for 15 minutes or longer.CU3 Initiating Condition:

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

3,4 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Voltage is < 115 VDC on required 125 VDC battery busses B and C for > 15 minutes.D No Change E Difference 1: Deviation 1) Listed site specific voltage and equipment to ensure timely classification.

2) Removed the word "indicated" this will allow for an indication problem to not cause confusion on the need to declare.Page 44 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CU5 Initiating Condition:

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

Cold Shutdown, Refueling, Defuled Example Emergency Action Levels: (1 or 2 or 3)1. Loss of ALL of the following onsite communication methods: (site-specific list of communications method 2. Loss of ALL of the following ORO communications s) methods: (site-specific list of communications methods)3. Loss of ALL of the following NRC communications methods: (site-specific list of communications methods)CU4 Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

3,4, D Emergency Action Level (EAL): 1. Loss of all Table C1 Onsite communications capability affecting the ability to perform routine operations.

OR 2. Loss of all Table CI Offsite communication capability affecting the ability to perform offsite notifications.

OR 3. Loss of all Table CI NRC communication capability affecting the ability to perform NRC notifications.

M No Change --- Difference I Deviation 1) Listed site specific communications methods to ensure timely classification Table Cl Communications Capability System Onsite Offsite NRC Plant Paging System X Station Radio X Conventional Telephone lines X X X Satellite Phone system X X NARS X HPN X X ENS X X Page 45 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CA3 Initiating Condition:

ALERT Inability to maintain the plant in cold shutdown.Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Alert promptly upon determining that the applicable has been exceeded, or will likely be exceeded.1. UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown temperature limit) for greater than the duration specified in the following table.2. UNPLANNED RCS pressure increase greater than (site-specific pressure reading). (This EAL does not apply during water-solid plant conditions.

[PWR])CA5 Initiating Condition:

Inability to maintain plant in cold shutdown.Operating Mode Applicability:

3,4 Emergency Action Levels (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.

1. UNPLANNED rise in RCS temperature

> 212 0 F for > Table C2 duration.OR 2. UNPLANNED RCS pressure rise > 10 psig as a result of temperature rise.Table C2 -RCS Heat-up Duration Thresholds H No Change E Difference 1 Deviation 1) Listed site specific Technical Specification cold shutdown temperature limit to ensure timely classification.

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

3) Added wording relating the temp and press rise to a loss of decay heat removal capability as per the developer notes for PWR's 4) Added the following note, taken from the basis section of the EAL, to bring it to the attention of the SM/ED when using the "procedure matrix" (1 lx17 quick reference control room document) "A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification." Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration Status Intact (but not RCS Reduced Not Applicable 60 minutes*Inventory

[PWR])Not Intact (or at Established 20 minutes*reduced inventory[PWR]) -Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

RCS Status Containment Closure Heat-up Duration Status Intact Not Applicable 60 minutes*Established 20 minutes*Not Intact Not Established 0 minutes If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, then EAL #1 is not aoelicable.

Page 46 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification CU3 Initiating Condition:

UNUSUAL EVENT UNPLANNED increase in RCS temperature.

Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.1. UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown temperature limit).2. Loss of ALL RCS temperature and (reactor vessel/RCS

[PWR] or RPV[BWRI) level indication for 15 minutes or longer.CU5 Initiating Condition:

UNPLANNED rise in RCS temperature.

Operating Mode Applicability:

3,4 Emergency Action Levels (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.

1. UNPLANNED rise in RCS temperature

> 212 0 F.OR 2. Loss of the following for > 15 minutes.* ALL RCS temperature indications AND* ALL RPV level indications H:1 No Change W Difference 1: Deviation 1) Listed site specific Technical Specification cold shutdown temperature limit to ensure timely classification.

2) Changed the word increase to rise in the initiating condition to be consistent with operations language and training.3) Added the following note, taken from the basis section of the EAL, to bring it to the attention of the SM/ED when using the "procedure matrix" (1 1x17 quick reference control room document) "A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification." Page 47 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification CGI Initiating Condition:

GENERAL EMERGENCY Loss of (reactor vessel/RCS

[PWRJ or RPV [BWR]) inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

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

[PWRJ or RPV [BWRJ) vessel level less than (site-specific level) for 30 minutes or longer.AND b. ANY indication from the Containment Challenge Table 2. a.. (Reactor vessel/RCS

[PWR] or RPV [BWR]) vessel level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following:

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

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

AND c. ANY indication from the Containment Challenge Table).CG6 Initiating Condition:

Loss of reactor vessel / RCS inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

3,4 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. a. RPV water level < 0 Inches TAF for > 30 minutes.AND b. Any Containment Challenge Indication (Table C4)OR 2. a. RPV water level cannot be monitored for > 30 minutes.AND b. Core uncovery is indicated by ANY of the following:

1. Table C3 indications of a sufficient magnitude to indicate core uncovery.OR 2. Refuel Floor Area Radiation Monitor C-10, North Wall, reading>3 R/hr.H-1 No Change W Difference

[:] Deviation 1) Listed site specific levels, radiation monitors, and sumps and tanks to ensure timely classification.

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

AND c. ANY Containment Challenge Indication (Table C4)Table C3 Indications of RCS Leakage Table: Containment Challenge Table* CONTAINMENT CLOSURE not established'

(Explosive mixture) exists inside containment

UNPLANNED increase in containment pressure* Secondary containment radiation monitor reading above (site-specific value) [BWR]'if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

Table C4 Containment Challenge Indications

Primary Containment Hydrogen Concentration

> 6% and Oxygen Concentration

> 5%* UNPLANNED rise in containment pressure* CONTAINMENT CLOSURE not established*

ANY Secondary Containment radiation monitor > EMG-3200.11I Maximum Safe (Table 12).* 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.________________________________________________

I] ________________________________________________

Page 48 of 67 NEI 99-01 Rev 6 Proposed EAL I Justification CS1 Initiating Condition:

SITE AREA EMERGENCY Loss of (reactor vessel/RCS

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

Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2 or 3)Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes time has been exceeded, or will likely be exceeded.1. a. CONTAINMENT CLOSURE not established.

AND b. (Reactor vessel/RCS

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

AND b. (Reactor vessel/RCS

[PWR] or RPV [BWR1) level less than (site-specific level).3. a. (Reactor vessel/RCS

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

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

[PWRJ* UNPLANNED increase in (site-specific sump and/or tank levels) of sufficient magnitude to indicate core uncovery* (Other site-specific indications)

CS6 Initiating Condition:

Loss of reactor vessel / RCS inventory affecting core decay heat removal capabilities.

Operating Mode Applicability:

3,4 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. With CONTAINMENT CLOSURE not established, RPV water level< 56 inches TAF.OR 2. With CONTAINMENT CLOSURE established, RPV water level< 0 inches TAF.OR 3. a. RPV water level cannot be monitored for > 30 minutes AND b. Core uncovery is indicated by ANY of the following:

Table C3 indications of a sufficient magnitude to indicate core uncovery.OR* Refuel Floor Area Radiation Monitor C-10, North Wall, reading>3 R/hr.Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

H No Change E Difference 1 Deviation 1) Listed site specific values for level, radiation monitors, and sumps and tanks to ensure timely classification.

Page 49 of 67 NEI 99-01 Rev 6 Proposed EAL Justification CA1 Initiating Condition:

ALERT Loss of (reactor vessel/RCS

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

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded.1. Loss of (reactor vessel/RCS

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

[PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer AND b. UNPLANNED increase in (site-specific sump and/or tank) levels due to a loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory.

CA6 Initiating Condition:

Loss of reactor vessel I RCS inventory Operating Mode Applicability:

4,5 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of RPV inventory as indicated by level < 86 inches TAF.OR 2. a. RPV water level cannot be monitored for > 15 minutes.AND b. Loss of RPV inventory per Table C3 indications.

D No Change E Difference I: Deviation 1) Listed site specific levels, and sumps and tanks to ensure timely classification.

Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

Page 50 of 67 NEI 99-01 Rev 6 1 Proposed EAL Justification Cul Initiating Condition:

UNUSUAL EVENT UNPLANNED loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory for 15 minutes or longer.Operating Mode Applicability:

Cold Shutdown, Refueling Example Emergency Action Levels: (1 or 2)Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.1. UNPLANNED loss of reactor coolant results in (reactor vessel/RCS

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

[PWR] or RPV [BWR]) level cannot be monitored.

AND b. UNPLANNED increase in (site-specific sump and/or tank) levels.CU6 Initiating Condition:

UNPLANNED loss of reactor vessel / RCS inventory for 15 minutes or longer.Operating Mode Applicability:

4,5 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. UNPLANNED loss of reactor coolant results in the inability to restore and maintain RPV level above the procedurally established lower limit for> 15 minutes.OR 2. a. RPV water level cannot be monitored AND b. Loss of RPV inventory per Table C3 indications.

D: No Change E Difference 1 Deviation 1) Described "a required lower limit" as a procedurally established lower limit, and listed site specific sumps and tanks to ensure timely classification.

Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

Page 51 of67 NEI 99-01 Rev 6 Proposed EAL Justification HG1 Initiating Condition:

GENERAL EMERGENCY HOSTILE ACTION resulting in loss of physical control of the facility.Operating Mode Applicability:

All Example Emergency Action Levels: 1. a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift supervision).

AND b. EITHER of the following:

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

Reactivity control Core cooling [PWR] / RPV water level [BWR]* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT HG1 Initiating Condition:

HOSTILE ACTION resulting in loss of physical control of the facility.Operating Mode Applicability:

1,2,3,4, D Emergency Action Level (EAL): 1. A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.AND 2. a. ANY Table H 1 safety function cannot be controlled or maintained.

OR b. Damage to spent fuel has occurred or is IMMINENT FD No Change I Difference FIDeviation

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

Table H1 Safety Functions* Reactivity Control (ability to shutdown the reactor and keep it shutdown)* RPV Water Level (ability to cool the core)* RCS Heat Removal (ability to maintain a heat sink)Page 52 of 67 NEI 99-01 Rev 6 Proposed EAL Justification HSI HSI Initiating Condition:

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

Initiating Condition:

HOSTILE ACTION within the Protected Area.Operating Mode Applicability:

1, 2,3,4, D No Change D Difference FIDeviation

1) List site security shift supervision as Security Force.All Example Emergency Action Levels: A HOSITLE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-security shift supervision).

Emergency Action Level (EAL): A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.Page 53 of 67 NEI 99-01 Rev 6 1 Proposed EAL I Justification HA1 Initiating Condition:

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

All Example Emergency Action Levels: (1 or 2)1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site-specific security shift supervision).

2. A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.HA1 Initiating Condition:

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

1,2,3,4, D Emergency Action Level (EAL): 1. A validated notification from NRC of an aircraft attack threat < 30 minutes from the site.OR 2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.W No Change FIDifference

[- Deviation 1) List site security shift supervision as Security Force.Page 54 of 67 NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENT Confirmed SECURITY CONDITION or threat.Operating Mode Applicability:

HUI HU1 Initiating Condition:

Confirmed SECURITY CONDITION or threat.Operating Mode Applicability:

D No Change M Difference F- Deviation 1) List site security shift supervision as Security Force.2) Further described credible security threat through listing a site specific procedure.

All Example Emergency Action Levels: (1 or 2 or 3)1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision).

2. Notification of a credible security threat directed at the site.3. A validated notification from the NRC providing information of an aircraft threat.1,2, 3,4, D Emergency Action Level (EAL): 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. 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 67 NEI 99-01 Rev 6 Proposed EAL I Justification HS6 Initiating Condition:

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

All Example Emergency Action Levels: (1 and 2)Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that (site-specific number of minutes) has been exceeded, or will likely be exceeded.1. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

2. Control of ANY of the following key safety functions is not reestablished within (site-specific number of minutes).Reactivity control* Core cooling [PWR] / RPV water level [BWR9* RCS heat removal HS2 Initiating Condition:

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

1,2,3,4, D Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1 A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per ABN-30 Control Room Evacuation.

AND 2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.D No Change M Difference FIDeviation

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

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 room still manned, due to testing or equipment failure.Table H1 Safety Functions* Reactivity Control (ability to shutdown the reactor and keep it shutdown)* RPV Water Level (ability to cool the core)* RCS Heat Removal (ability to maintain a heat sink)Page 56 of 67 NEI 99-01 Rev 6 Proposed EAL Justification HA6 Initiating Condition:

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

Operating Mode Applicability:

All Example Emergency Action Levels: An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

HA2 Initiating Condition:

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

Operating Mode Applicability:

1,2, 3,4, D Emergency Action Level (EAL): A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per ABN-30 Control Room Evacuation.

[D No Change F Difference

[- Deviation 1) EAL uses the site specific Control Room evacuation procedures to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedures the Control Room would enter should such an event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate shutdown were to be listed, 2) Changed "An event" to" A Control Room evacuation" to remove confusion if partial plant control was transferred to outside the control room with the control room still manned, due to testing or equipment failure.Page 57 of 67 NEI 99-01 Rev 6 Proposed EAL [ Justification HU4 HU3 Initiating Condition:

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

Initiating Condition:

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

H-- No Change W- Difference F Deviation 1) Listed site specific list of plant rooms or areas that contain SAFETY SYSTEM equipment to ensure timely classification.

2) The site ISFSI is located within the Plant Protected Area and as such is not specifically addressed in EAL #3 and #4. This is in accordance with the EAL developer notes.3) Added additional note to easily direct the operator to escalation ICs All Example Emergency Action Levels: (1 or 2 or 3 or 4)Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

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

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

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

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

1,2,3,4, D Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* Escalation of the emergency classification level would be via IC CA2 or MA5 1. A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of the following FIRE detection indications:

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

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

Field verification of a single fire alarm Table H2 Vital Areas* Reactor Building (when inerted the Drywell is exempt)* 4160V Switchgear Rooms (1C and 1D)* Control Room Complex (MOB, Upper and Lower Cable Spreading Rooms)* Main Transformer/Condensate Transfer Pad* Intake Structure* #1 EDG Vault* #2 EDG Vault* EDG Fuel Oil Storage Tank OR 2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of a FIRE).AND b. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.OR 3. A FIRE within the plant PROTECTED AREA not extinguished in< 60 minutes of the initial report, alarm or indication.

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

Page 58 of 67 NEI 99-01 Rev 6 Proposed EAL I Justification Initiating Condition:

UNUSUAL EVENT Seismic event greater than OBE levels.Operating Mode Applicability:

HU'All Example Emergency Action Levels: Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: a. (site-specific indication that a seismic event met or exceeded OBE limits)HU4 Initiating Condition:

Seismic event greater than OBE levels.Operating Mode Applicability:

1,2,3,4, D Emergency Action Level (EAL): Note: Escalation of the emergency classification level would be via IC CA2 or MA5 For emergency classification if EAL 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in < 15 mins of the event.Seismic event as indicated by: 1. Control Room personnel feel an actual or potential seismic event.AND 2. ANY one of the following confirmed in < 15 mins of the event: " The earthquake resulted in Modified Mercalli Intensity (MMI) _ VI and occurred < 3.5 miles of the plant." The earthquake was magnitude

> 6.0" The earthquake was magnitude

> 5.0 and occurred < 125 miles of the plant.D No Change M Difference F Deviation 1) Used Alternate developer notes allowed wording since specific Control Room indication of a seismic event > OBE is not available.

2) Added additional compensatory thresholds based on the guidance provided in US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions Appendix A, also added a 15 min time limit to ensure timely classification when these additional thresholds may not be readily available.

3) Added additional notes to easily direct the operator to the escalation ICs, as well as to ensure if unable to determine size of earthquake lAW 2.b then the SM / ED will determine classification within 15 mins of the event Page 59 of 67 NEI 99-01 Rev 6 HA5 Initiating Condition:

ALERT Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.Operating Mode Applicability:

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

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

AND b. Entry into the room or area is prohibited or impeded.Proposed EAL Justification

-4 i HA5 Initiating Condition:

Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.Operating Mode Applicability:

3,4 Emergency Action Level (EAL): Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted.

1. Release of a toxic, corrosive, asphyxiant or flammable gas in a Table H3 area.Table H3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building 51' elevation Modes 3 and 4 AND 2. Entry into the room or area is prohibited or impeded D No Change I Difference r-1 Deviation 1) Listed plant specific rooms and areas with entry related mode applicability to ensure timely classification.

2) The Operating Mode Applicability of this EAL has been revised from All Modes to modes 3 and 4 due to the mode applicability of the areas of concern in Table H-3.Page 60 of 67 NEI 99-01 Rev 6 Proposed EAL Justification HU3 Initiating Condition:

UNUSUAL EVENT Hazardous Event Operating Mode Applicability:

All Example Emergency Action Levels: (1 or 2 or 3 or 4)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 or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).4. A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.5. (Site-specific list of natural or technological hazard events)HU6 Initiating Condition:

Hazardous Event Operating Mode Applicability:

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

Escalation of the emergency classification level would be via IC CA2 or MA5 1. Tornado strike within the PROTECTED AREA.OR 2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode.OR 3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).OR 4. A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.OR 5. Abnormal Intake Structure level, as indicated by EITHER:* > 4.5 ft. MSL (> 4.25 psig on PI-533-1172 and PI-533-1173 or > 4.5 ft MSL on CR-423-11 pt 24 and pt 23).OR* b. < -3.0 ft. MSL (L 0.95 psig on PI-533-1172 and PI-533-1173 or < -3.0 ft MSL on CR-423-11 pt 24 and pt 23).MSL = Mean Sea Level[- No Change FIDifference

[- Deviation 1) Included River water level as part of the site specific list of natural or technological hazard events. The EAL values selected are the current Approved UE EAL values.2) Changed the word "needed" to "required by Technical Specifications" in the EAL to be consistent with terminology used by operators and minimize confusion.

3) Added additional note to easily direct the operator to escalation ICs Page 61 of 67 NEI 99-01 Rev 6 Proposed EAL Justification HG8 HG7 Initiating Condition:

GENERAL EMERGENCY Initiating Condition:

No Change EIDifference FD eviation Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.

Operating Mode Applicability:

All Example Emergency Action Levels: 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 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 the facility.

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

Operating Mode Applicability:

1, 2, 3,4, D Emergency Action Level (EAL): Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT 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 the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area..Page 62 of 67 NEI 99-01 Rev 6 1 Proposed EAL .I Justification HS8 HS7 Initiating Condition:

SITE AREA EMERGENCY Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.

Operating Mode Applicability:

All Example Emergency Action Levels: Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that 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 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 Guideline exposure levels beyond the site boundary.Initiating Condition:

M No Change I Difference

[IDeviation Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.

Operating Mode Applicability:

1,2,3,4, D Emergency Action Level (EAL): Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major 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 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 expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.Page 63 of 67 NEI 99-01 Rev 6 Proposed EAL Justification HA6 HA7 Initiating Condition:

ALERT Initiating Condition:

FW No Change 1 Difference I Deviation Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.Operating Mode Applicability:

All Example Emergency Action Levels: 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 substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.Operating Mode Applicability:

1,2,3,4, D Emergency Action Level (EAL): Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.Page 64 of 67 NEI 99-01 Rev 6 [ Proposed EAL [ Justification HU7 Initiating Condition:

UNUSUAL EVENT Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT.Operating Mode Applicability:

All Example Emergency Action Levels: Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.HU7 Initiating Condition:

Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT.Operating Mode Applicability:

1,2,3,4, D Emergency Action Level (EAL): Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.W No Change ODDifference F-1 Deviation Page 65 of 67 NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition:

UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY.Operating Mode Applicability:

E-HUI 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.E-HUI Initiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY.Operating Mode Applicability:

1,2, 3,4, D Emergency Action Level (EAL): Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading:* > 800mr/hr (gamma + neutron) 3 feet from the HSM surface OR* > 200mr/hr (gamma + neutron) outside the HSM door on centerline of DSC OR* > 40mr/hr (gamma + neutron) end of shield wall exterior D 1 No Change -1Difference 1 Deviation 1) Listed 2x the site specific cask specific allowable radiation level as per Oyster Creek Generating Station ISFSI 10CFR72.212 Evaluation Rev. 5 2) Not all technical specification radiation readings were on contact, one is a radiation reading at 3ft required by technical specification.

Modified the EAL by removing the "on the surface" requirement, and required "a radiation reading" allowing for the technical specification 3ft reading to be added to the EAL.Page 66 of 67 N'C C N'C 0) 4 Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RG1l Initiating Condition:

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

~ ~ ~ ~ ~ ~ ~ ~~ ~~.. .. .i ..... .........

S:: Operating Mpde Applicacbility 1, 2,3,4, D.Emergenrcy Action Level (EAL): Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time.* 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. 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. >10O0mRemTEDE OR b. > 5000 mRem CDE Thyroid OR 3. Field survey results at or beyond the site boundary indicate EITHER: a. Gamma (closed window) dose rates >1000 mR/hr are expected to continue for > 60 minutes.OR b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

xxxMonth2015 OCGS 3-33 EP-AA-1010 Rev XX (I afn (rga rUanni --l * .* n U4 *r~n *. .nv I=yalnn NJrlonr A~,@4car t~rnaI, (~n~r~tinn

~t2tir~n Ann~v Fv~Inn Mm~r~I~2r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG1 (cont),Emergency Acon -Level (EAL) (cont): -gen.......

y. A ction... ...... .../Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency Main Stack RAGEMS 8.69 E+01 uCi/cc HRM OR 3.53 E-08 amps HRM Turbine Building RAGEMS 7.17 E-01 uCi/cc HRM HRM = High Range Monitor B a s is :.. ..... .... ... .....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 mRem while the 5000 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Basis 1.2.3.4.5.6.Reference(s)::

NEI 99-01 Rev 6, AG1 EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent Oyster Creek Generating Station BNE Correspondence dated February 1, 2007 CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek EP-AA-1 10-200, Dose Assessment EP-AA-1 10-201, On Shift Dose Assessment EAL Threshold xxxMonth2015 OCGS 3-34 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex I=xelon Nuclear Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RS1 Initiating Condition:

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

1,2,3,4, D Emergencyi Action ..Leel (EAL)...........:.

Notes: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time." 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. 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. > 100 mRem TEDE OR b. > 500 mRem CDE Thyroid OR 3. Field survey results at or beyond the site boundary indicate EITHER: a. Gamma (closed window) dose rates >100 mR/hr are expected to continue for > 60 minutes.OR b. Analyses of field survey samples indicate > 500 mRem CDE-Thyroid for 60 minutes of inhalation.

xxxMonth2015 OCGS 3-35 EP-AA-1010 Rev XX IFyolnn Nma~lr~I~Ouster Creek Generatinn Station nnex Fvalnn N"rlanr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS1 (cont)Emergency

Acti6, evel (EAL) (cont): Table RI Effluent Monitor Thresholds Effluent Monitor Site Area Emergency Main Stack RAGEMS 8.69 E+00 uCi/cc HRM OR 3.53 E-09 amps HRM Turbine Building RAGEMS 3.65 E+05 cpm LRM HRM = High Range Monitor LRM = Low Range Monitor 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 mRem while the 500 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Escalation of the emergency classification level would be via IC RGI.Basis Ref.ere..nce(s):.

." .". ..1. NEI 99-01 Rev 6, AS1 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek 5. EP-AA-1 10-200, Dose Assessment

6. EP-AA-1 10-201, On Shift Dose Assessment xxxMonth2015 OCGS 3-36 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex...... ....Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA1.Initiating Condition:

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

1,2,3,4, D ,Emergency Action Level (EAL): ..... ..... .... .......Notes: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time." 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. 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 xxxMonth2015 OCGS 3-37 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT OR xxxMonth2015 OCGS 3-38 EP-AA-1010 Rev XX n afar (rg%,mL- ftanarnfin Q+!a+irn Annnv 9=vd%1tn Kiiii-Idmor z~a4~ U~r~ ~nr+ r 4~i~ nnVf Vd IL**i t4L **I.E f l ****t.*S Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RAI (cont)Emergency,, A-ction, Leve! (EAL) (q6nt)-: 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 a 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 Main Stack RAGEMS 8.69 E-01 uCi/cc HRM OR 3.53 E-10 amps HRM Turbine Building RAGEMS 3.65 E+04 cpm LRM HRM = High Range Monitor LRM = Low Range Monitor 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 mRem while the 50 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.The radwaste liquid discharge system is currently closed off with a plant modification installed blank flange. To perform a discharge would require a plant modification to remove the flange. Since the liquid radwaste system is not operable, no EAL threshold has been developed for this release point.xxxMonth2015 OCGS 3-39 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT Escalation of the emergency classification level would be via IC RS1.RAI (cont)."Basis Reference(s):

1. NEI 99-01 Rev 6, AA1 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek 5. ABN-27, Inadvertent Overboard Radioactive Release or Cross Contamination

6. EP-EAL-0617, Oyster Creek Criteria for Choosing Liquid Effluent EAL Threshold Values xxxMonth2015 OCGS 3-40 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU1 I1nitiating Condition:.

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating:

Mode. Applicability:_

1,2,3,4, D Emergency Action Level (EAL): Notes:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time.* 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. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 60 minutes: Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Event Main Stack RAGEMS 4.07 E+03 cps LRM Turbine Building RAGEMS 4.16 E+02 cpm LRM LRM = Low Range Monitor OR 2. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.xxxMonth2015 OCGS 3-41 EP-AA-1 010 Rev XX Oyster Creek Generatinq Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RUI (cont)Bsis: 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.

Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.The radwaste liquid discharge system is currently closed off with a plant modification installed blank flange. To perform a discharge would require a plant modification to remove the flange. Since the liquid radwaste system is not operable, no EAL threshold has been developed for this release point.EAL #1 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored-gaseous effluent pathways.EAL #2 Basis 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 RAI.xxxMonth2015 OCGS 3-42 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RU1 (cont)Basis Reference(s):

.1. NEI 99-01 Rev 6, AU1 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek 5. ABN-27, Inadvertent Overboard Radioactive Release or Cross Contamination xxxMonth2015 OCGS 3-43 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annexm m Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RG2'Initiating Condition:

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

1,2,3,4, D ,Emergency Action Level (EAL): Note: The Emergency Director should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for 60 minutes or longer.* : ... ... ........:>. .... .... 7 ...... .... ..... .........,., ... .... ... ..... ... .......Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

... ... .~ ~ ~ ~ ~ ~ ~ .. ... .. .. ..... ....... ..........

..... ..... ... ... ... ..... ..' .. .... ... .:..... ... " .... ........Bas.is Reference(s):

1. NEI 99-01 Rev 6, AG2 xxxMonth2015 OCGS3-44 EP-AA-1010 Rev XX Ovstar Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS2 Initiating Condition:

-Spent fuel pool level at (site-specific Level 3 description).

0.erating -Mode .Applicability:

1,2,3,4, D'Emergency Actior Levyel" (EAL)i:,, Lowering of spent fuel pool level to (site-specific Level 3 value).Basis:.. ...This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC RG1 or RG2.Basis Reference(s):i

1. NEI 99-01 Rev 6, AS2 xxxMonth2015 OCGS 3-45 EP-AA-1010 Rev XX v I au r rrolk 11Inarninn I tntinn AnnIl V ExVlon Nuc*lelr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA2 Initiating Condition:

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

.................

1,2,3,4, D.Emergency, Action, Level (EAL):-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 mRem/hr.OR 3. Lowering of spent fuel pool level to (site specific Level 2 value).Table R2 Refuel Floor ARM's* C-5, Crit Mon* C-10, North Wall* C-9, North Wall" B-9, Open Floor REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange..IMMINENT:

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

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

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

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.xxxMonth2015 OCGS3-46 EP-AA-1010 Rev XX

() Qfar ('raak rZanarnfin Afnfinn Annoy , I=xAInn

(, aIInnIUIIII~l

¥taUiUU Annv Fx Inn Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA2 (cont)Basis (cont):.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.EAL #1 Basis This EAL escalates from RU2 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 a rise 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 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).

EAL #3 Basis: Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency would be based on either Recognition Category R or C ICs.xxxMonth2015 OCGS3-47 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 -OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA2 (cont)(Basis Refe-rence~s):

-:" .1. NEI 99-01 Rev 6, AA2 2. RAP G-7-a, SKM SRG TNK LVL LO-LO 3. RAP-10F-1-m, Crit Mon C5 Hi 4. RAP-1 OF-3-m, North Wall C9 Hi Vent Trip 5. RAP-10F-2-m, North Wall 010 Hi 6. RAP-1 OF-4-m, North Wall B9 Hi Vent Trip xxxMonth2015 OCGS3-48 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU2 Condition:

i_~ ~::iii: ,, ii i " .--_UNPLANNED loss of water level above irradiated fuel..operating ModeApplicability:

-, 1,2,3,4, D Emergency Action Level (EAL): 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following: " Refueling Cavity water level < 583 inches (GEMAC Wide Range, floodup calibration).

OR* Indication or report of a drop in water level in the REFUELING PATHWAY.AND b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R2.Table R2 Refuel Floor ARM's* C-5, Crit Mon* C-10, North Wall" C-9, North Wall* B-9, Open Floor xxxMonth2015 OCGS 3-49 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RU2 (cont)UNPLANNED:

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

The cause of the parameter change or event may be known or unknown.REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.This IC addresses a loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level 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 a rise 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 rise 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 RA2.Basis Rfrence(sý)':

1. NEI 99-01 Rev 6, AU2 2. RP-AA-203 Exposure Control and Authorization

3. RAP-G-7-a, SKM SRG TNK LVL LO-LO 4. 205.94.0 RPV Floodup Using Core Spray 5. 205.95.0 Reactor Flood-up / Drain-down

6. FSAR Figure 7.6-3 xxxMonth2015 OCGS 3-50 EP-AA-1010 Rev XX n *for ('rdmdak ('Zonarnfin Afnfinn nnayr A~,etar (r~at t~a~r~tnn

~~tin Annv FvInn MNur-Io2r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA3 initiating Condition:

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

1,2,3,4, D Em ergency A ction Level (EA L): ............

... .. ...........

.............

...Note: If the equipment in the 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. Dose rate > 15 mR/hr in ANY of the following Table R3 areas: Table R3 Areas Requiring Continuous Occupancy* Main Control Room" Central Alarm Station -(by survey)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 51' elevation*

Modes 3 and 4*Areas required to establish shutdoWn GOGiii xxxMonth2015 OCGS 3-51 -EP-AA-1010 Rev XX fafga4r (f rgagl a~narisfin -Q+m4ie-n An ngv IPvftln NKl,',dar z V * -*- l *L **i L *r,***flt

* -*A Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 (cont)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.

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.

Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR, therefore Table R4 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in service. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding beyond that required by procedures, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).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 rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.xxxMonth2015 OCGS 3-52 EP-AA-1010 Rev XX Ovster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA3 (cont).Basis (cont):* 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 R, C or F ICs.Basis Reference(s):

1. NEI 99-01 Rev 6, AA3 2. ABN-29, Plant Fires 3. EMG-3200.1 1, Secondary Containment Control Safe Shutdown Area xxxMonth2015 OCGS 3-53 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RU3 lnitiating Condition:

.......Reactor coolant activity greater than Technical Specification allowable limits.* ...... ..> ........ .. ... ...... ......................

..... ........ * ............ ... .. ... .... ... .... .. ..... ........0.perating Mode Applicabillity-1,2 Emergency Action Level (EAL): 1. Offgas system radiation monitor Hi-Hi alarm.OR 2. Specific coolant activity > 4.0 uCllgm Dose equivalent 1-131.Basis: .........

..,;.This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.

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

An Unusual Event is onlY warranted when actlual fue'Pl Glad 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 un~les another cau.e knoWR.Escalation of the emergency classification level would be via ICs FAl or the Recognition Category R ICs.Basis Referenc-e(s):

........ ......... .1. NEI 99-01 Rev 6, SU3 2. Technical Specifications 3.6.A 3. ABN-26, High Main Steam Line or Off Gas Activity 4. RAP10F-1-c, Offgas HI-HI xxxMonth2015 OCGS 3-54 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 In itiating C ond itio n: ... ..... ..., ... .. ... ... ..... .... ....Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.Operating Mode Applicability:

1,2..... .... ....... ...... .... ...... ...... ....... ... ..... ...........

.............

.. .: ' .... ..:....... ..: ... ....ii :! : :, ..Emnerge~ncy Action Le6vel (EAL)ý: 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 9 Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-55 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FSl Initiating Condi~tion.:

'...- -.-Loss or Potential Loss of ANY two barriers.OP*6eratifngModie Applicability:

... ..... .... .1,2 Emergency ActionmLevel (EAL): Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.B a s i :: ..........

: ' .:":::i~ i,:i~ii' : " ..... :: .........

.": : ..... ..* ......... .... .... ........ ....ýBasis:, 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):I

..6, .9 1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-56 EP-AA-1 010 Rev XX Ovstar Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiatin g Con;ditio-n":'

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

1,2 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-2 xxxMonth2015 OCGS 3-57 EP-AA-1010 Rev XX C)v4mtPrCrcek C' onerating Station Annexr flv~tr Cr~~k (~n~r~inn

~~tinnAnnEvFvlIn Nuclea~ar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FCl Initiating C ondition:

.......................

..........RCS Activity Operatig Mode Applicability:

...........1,2'Fission Product Barrier (FPB)Threshoidi:

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

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

There is no Potential Loss threshold associated with RCS Activity.Basis Reference(s)y:

......1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-58 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex IFy,-*le~n KimJlialr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 ntiating Condition:, -RPV Water Level Oprating MdAplcability:

1,2 Fission..

Pro ct Barrier (FPB).Threshold:

.LOSS 1. Plant conditions indicate primary containment flooding is required.POTENTIAL LOSS 2. RPV water level cannot be restored and maintained

> 0 inches TAF.OR 3. RPV water level cannot be determined.

Basis: Loss Threshold

1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.This is identified in the BWROG EPGs/SAGs when the phrase, "Primary Containment Flooding Is Required," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "Primary containment flooding required," also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.

Potential Loss Threshold

2 and #3 Basis This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.The RPV water level threshold is the same as RCS Barrier RC2 Loss threshold.

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

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

EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below xxxMonth2015 OCGS 3-59 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)B asis (cont):, ... ......the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.

For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

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

Basis R eference (s):;,,%, .......... .. ... ..........

.. .1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.01A, RPV Control -No ATWS 3. EMG-3200.01 B, RPV Control -With ATWS 4. EMG-3200.08A, RPV Flooding -No ATWS 5. EMG-3200.08B, RPV Flooding -With ATWS 6. EMG-3200.02, Primary Containment Control xxxMonth2015 OCGS 3-60 EP-AA-1010 Rev XX Oyster Creek Generatinq Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:

.......... ...... ... .Primary Containment Radiation 0.petn .Moe Applicability 1,2 Fission product Barrieri (FPB) Thresholdi:

LOSS Containment Hi Range Radiation Monitoring System (CHRRMS) reading > 530 R/hr.Ba'sis:.The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 iCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Barrier RC5 Loss Threshold 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.

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

Basis Reference(s):

".: ... ... ...... .. ... " .=: i " ........ : == = ° : '" 1. NEI 99-01 Rev 6, Table 9-F-2 2. Core Damage Assessment Methodology xxxMonth2015 OCGS 3-61 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating CIo'nditi~oln-:,,..................

Emergency Director Judgment.Operating Mode Applicability:

..........

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

LOSS 1. Any condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.POTENTIAL LOSS 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.Basis:... is .: ... .. ....; ; ... *. .. ...: ' ...' ' : : :. ..Loss Threshold

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

2 Basis 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.

.~ ~ ~~~~ ~ ~~~~~~~~~...

...... .. ...... ... .. .. ., *.. ..-: ...... _ -.! ." ....Bas is Reference(s):el

1. NEI 99-0 1 Rev 6, Table 9-F-2 xxxIVIonth2015 OCGS 3-62 EP-AA-1 010 Rev XX C) *for~ *'oo S%.* Zoani Afnfinn A*S IPvalrn I~elnrlr A~,et~r t~raaI, (~an~r2tinn

~t2tinn Ann~v Fv~Inn M..r~I~2r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating, Condition:

RPV Water Level Operating Mode Applicability:

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

-, LOSS 1. RPV water level cannot be restored and maintained

> 0 inches TAF.OR 2. RPV water level cannot be determined.

B asis: ... ..... ......This water level corresponds to the Top of Active Fuel (TAF) and is used in the EOPs to indicate challenge to core cooling.The RPV water level threshold is the same as Fuel Clad Barrier FC2 Potential Loss threshold.

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

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.xxxMonth2015 OCGS 3-63 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex IPvalnn Mir~lanr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)Basis ....o.nt..: In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.

For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

There is no RCS Potential Loss threshold associated with RPV water level.Basis Reference-(s):

1. NEI 99-01 Rev 6, Table 9-F-2 2. 2000-GLN-3200.01, Plant Specific Technical Guideline 3. 2000-BAS-3200.02, EOP Users.Guide xxxMonth2015 OCGS 3-64 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Pwi=Ynn M, r~lar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 in itia tin g .~i i n ... ..... ... ...... ... .. .

.... .... ... R, ... C 3 IIT fil Initiating Condition:

Primary Containment Pressure 0.perating Mode Applicability:

1,2*PFission Product Barrier (FPB) Threshold" LOSS 1. Drywell pressure > 3.0 psig.AND 2. Drywell pressure rise is due to RCS leakage Basis: ......... .... .. ........ ........ .... .. ... ........, ., ...""" ..The > 3.0 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure.

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

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

A stuck-open Electromatic Relief Valve (EMRV) or EMRV leakage is not considered either identified or unidentified leakage by Technical Specification and, therefore, is not applicable to this EAL.There is no Potential Loss threshold associated with Primary Containment Pressure.Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.01A, RPV Control -No ATWS 3. EMG-3200.02, Primary Containment Control 4. 2000-BAS-3200.02, EOP User's Guide xxxMonth2015 OCGS 3-65 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiat. g C:............i.n..

...... ....RCS Leak Rate Ope.ating Mode Applicability:

p ...... ... ..... , .... .... ... *.. .*.. ...... , .;, 1,2 Fission Product Barrier (FPB) Threshold-.-

LOSS 1. UNISOLABLE Main Steam Line (MSL), Isolation Condenser, Feedwater, or RWCU line break.OR 2. Emergency RPV Depressurization is required.POTENTIAL LOSS 3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature

> EMG-3200.11 Max Normal (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11 Max Normal (Table 12) operating level....... ..Basis:-UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Classification of a system break over system leakage is based on information available to the Control Room from the event. Indications that should be considered are:* Reports describing magnitude of steam or water release.* Use of system high flow alarms / indications, if available,* Significant changes in makeup requirements,* Abnormal reactor water level changes in response to the event.The use of the above indications provides the Control Room the bases to determine that the ongoing event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.xxxMonth2015 OCGS 3-66 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont).Basis:(cont):

Loss Threshold

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

If it is determined that the ruptured line cannot be promptly isolated, the RCS barrier Loss threshold is met.Loss Threshold

2 Basis Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open Electromatic relief valves (EMRVs) and keep them open. Even though the RCS is being vented into the torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.Potential Loss Threshold

3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as Trunnion room, Isolation Condenser, RWCU etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification.

A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials.

Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold#1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.xxxMonth2015 OCGS 3-67 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.1 1, Secondary Containment Control 3. 2000-GLN-3200.01, Plant Specific Technical Guideline xxxMonth2015 OCGS 3-68 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5.Initiating Condition:

...ia in C n i i n ..............

....... .........

......... .. .= = .. ......... =Primary Containment Radiation Operating Mode Applicability:

1,2 Fission Pro duc Barrier (FPB) Threshold':

LOSS Containment Hi Range Radiation Monitoring System (CHRRMS) reading > 10OR/hr......... .. .... .. .... .. ... ... ......... .. , .. .. ............._.. .. ...........

... ...*... ..*.... ..."" .... ..... .... .. .. .Basis:, The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier FC5 Loss Threshold since it indicates a loss of the RCS Barrier only.There is no Reactor Coolant System Potential Loss threshold associated with Primary Containment Radiation..Basis Reference(s):

--1 1. NEI 99-01 Rev 6, Table 9-F-2 2. EP-EAL-061 1, Criteria for Choosing Containment Radiation Monitor Reading Indicative of Loss of RCS Barrier xxxMonth2015 OCGS 3-69 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:

Emergency Director Judgment.0.perating Mode Applicability:.....

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

LOSS 1. Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.POTENTIAL LOSS 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.basis:,. -Loss Threshold

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

2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

.1. NEI 99-01 Rev 6, Table 9-F-2 xxxIVIonth2015 OCGS 3-70 EP-AA-1010 Rev XX Nuclear ()uctar Crrook Generating St~ation Annex Exelnn Nucleasr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:

..........

.....RPV Water Level O.perating Mode_ Applicability:

......... ...........

......... ...... .1,2 Fission Product Barrier (FPB) Threshold:'

" , ..-POTENTIAL LOSS Plant conditions indicate primary containment flooding is required.Basiis:-The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV water level. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be .restored and maintained and that core damage is possible.

BWR EOPs/SAMGs specify the conditions that require primary containment flooding.

When primary containment flooding is required, the EOPs are exited and SAMGs are entered. Entry into SAMGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

ýBasis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.01 B, RPV Control -With ATWS 3. EMG-3200.08A, RPV Flooding -No ATWS 4. EMG-3200.08B, RPV Flooding -With ATWS 5. EMG-3200.02, Primary Containment Control 6. EMG-3200.01A, RPV Control -No ATWS xxxlVIonth2015 OCGS 3-71 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Primary Containment Conditions OperatingMode Appiicability: , , 1,2 Fission Product Barrier (FPB) Threshold:

LOSS 1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.OR 2. Drywell pressure response not consistent with LOCA conditions.

POTENTIAL LOSS 3. Drywell pressure > 44 psig and rising.OR 4. a. Drywell or Torus Hydrogen concentration

> 6%.AND b. Drywell or Torus Oxygen concentration

> 5%.OR 5. Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded.:B a-isi ..............

.................

.........

................

..... ............................" ....................

.. ... r. .... ...... .. ... ........ ...:! ,, .. .. .. .. ... ..... .. ... ....UNPLANNED:

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

The cause of the parameter change or event may be known or unknown.Loss Threshold

1 and #2 Basis Rapid UNPLANNED loss of Drywell pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of Drywell integrity.

Drywell pressure should rise as a result of mass and energy release into the Drywell from a LOCA. Thus, Drywell pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned.

The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

A pressure suppression bypass path would not be an indication of a containment breach.xxxMonth2015 OCGS 3-72 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex P::vgm-In M, ,rl-ntr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)B a s is (c o n t) : ......... ... ..... ... ...Potential Loss Threshold

3 Basis The threshold pressure is the Drywell internal design pressure.

Structural acceptance testing demonstrates the capability of the Drywell to resist pressures greater than the internal design pressure.

A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.Potential Loss Threshold

4 Basis If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.Potential Loss Threshold

5 Basis The HCTL is a function of RPV pressure, torus temperature and torus water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

ýBasis Reference(s):.

1. NEI 99-01 Rev 6, Table 9-F-2 2. FSAR Update 6.2.1.1.3 3. Technical Specifications

5.2 Basis

4. EMG-3200.02 Primary Containment Control xxxMonth2015 OCGS 3-73 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating C ond itio n: ..........

.. .... ........ ..... ..Primary Containment Radiation:O perating M ode Applicability:

-- -...... ........ ........1,2 Fission Produt Barrier (FPB)Thres hold: POTENTIAL LOSS Containment Hi Range Radiation Monitoring System (CHRRMS) reading > 1210 R/hr.B asIs:' * -.. ......There is no Loss threshold associated with Primary Containment Radiation.

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, 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-2 2. Core Damage Assessment Methodology (CDAM)xxxMonth2015 OCGS 3-74 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6.. .* .,: ....... : , : : -v : : ....." :: .... ... ...... .... :' = i J Y ", initiating Condition:

...........

Primary Containment Isolation Failure.Operating Mode Applicability:

..1,2 Fission Product Barrier (FPB) "Threshold:

LOSS 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.OR 2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

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

a. Secondary Containment area temperature

> EMG-3200.11 Max Safe (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11 Max Safe (Table 12) operating level..Basis: ... ..... .UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss Threshold

1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment.

Examples include unisolable Main Steamline, Isolation Condenser line breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.xxxMonth2015 OCGS 3-75 EP-AA-1 010 Rev XX flv~t~r C.r0~k (~n0r2tinn

~t2tinn Ann~v Qu--tar Creak Generating Station nnox Evoinn Niurlanr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)Basis ( cont): .... ...,."...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.Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components.

Minor releases may also occur if a primary containment isolation valve(s)fails to close but the primary containment atmosphere escapes to an enclosed system.These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.Loss Threshold

2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded.

Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.

Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the Drywell high pressure scram setpoint) does not meet the threshold condition.

Loss Threshold

3 Basis The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.

EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials.

Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.xxxMonth2015 OCGS 3-76 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont).... * .... ..... .. ..... .. * ....... ...... .... .... ............ ... ..." .............

... .. .Basis(o In combination with RCS Barrier RC4 Potential Loss Threshold

3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.Basis R.eference(s):

.. .. .. ......: ....... ...... ...1. NEI 99-01 Rev 6, Table 9-F-2 2. 2000-GLN-3200.01, Plant Specific Technical Guideline 3. EMG-3200.02, Primary Containment Control 4. Support Procedures

-32, -34, -41, -44 5. 2000-GLN-3200.03, OCNGS Plant Specific Technical Guidelines for Severe Accident Guidelines

6. EMG-3200.1 1, Secondary Containment Control xxxMonth2015 OCGS 3-77 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Condition:

... .... .Emergency Director Judgment.Operating Mode Applicability:

1,2.Fission Product Barrier (FPB)_ Threshold:

LOSS 1. Any condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.POTENTIAL LOSS 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.B a s is :. ......... ...... .. .. ..Loss Threshold

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

2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

'Basis Referein.ce(s):.

1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-78 EP-AA-1 010 Rev XX lFv~lnn N. jlnnr Ov--tar Creek Generatinn Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 nitiating Condition:

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

.... .. .. ..!= : .. ..... .. ..... ...........

..... .' _ ...= ! : .: : ......... ... .. ....1,2:Emergency Action Level (EAL):...

J'...................

Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1 Loss of ALL offsite AC power to 4160V Buses 1C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D.AND 3. EITHER of the following:

a. Restoration of at least one 4160V Bus (1 C or 1 D) in < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is not likely.OR b. RPV water level cannot be restored and maintained

> -20 inches TAF.s :.= .... ..... -.- --... ...... ... s ..: : .. ... .. .. ........ .. ..... ...... 11 '.. ... ... ..... .... .. ........ ....... .. .. .... ... .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 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 any fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and xxxMonth2015 OCGS 3-79 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)Basis (cont):, event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation.

Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.:BaSiS Refer~ence(s):

...... : 1. NEI 99-01 Rev 6, SG1 2. UFSAR Section 8.2, Offsite Power System 3. ABN-37, Station Blackout 4. ABN-60, Grid Emergency 5. Regulatory Guide 1.155, Station Blackout 6. TDR-1099, "Station Blackout Evaluation Report" 7. 2000-BAS-3200.02, EOP User's Guide 8. 2000-GLN-3200.01, Plant Specific Technical Guideline 9. OCNGS Drawing BR 3000 10. ABN-36, Loss of Off-Site Power xxxMonth2015 OCGS 3-80 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 Initiating Condition:.

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

1,2 Emergency Action Level (EAL)' ..... ..... ...... .... .... .Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of ALL offsite AC Power to 4160V Buses 1 C and 1 D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D.AND 3. Failure to restore power to at least one 4160V Bus (1C or 1 D) in < 15 minutes from the time of loss of both offsite and onsite AC power.Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs RG1, FG1, MG1, or MG2.xxxMonth2015 OCGS 3-81 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)Basis 1.2.3.4.5.6.Reference(s):

NEI 99-01 Rev 6, SS1 UFSAR Section 8.2, Offsite Power System OCNGS Drawing BR 3000 ABN-36, Loss of Off-Site Power ABN-37, Station Blackout ABN-60, Grid Emergency xxxlVlonth2015 OCGS 3-82 EP-AA-1010 Rev XX I=xelon Nuclear COvster Creeck (GAneratincn Station Annex Exelon Nucleanr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAI Iintiating, Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.,pe ratingaMode Applicability:

1,2 Emergency Action Level. (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. AC power capability to 4160V Buses 1C and 1 D reduced to only one of the following power sources for > 15 minutes.* Startup Transformer SA* Startup Transformer SB* EDG-1 Emergency Diesel Generator* EDG-2 Emergency Diesel Generator AND 2. ANY additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including, the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

This IC provides an escalation path from IC MUI.An "AC power source" is a source recognized in ABNs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

xxxMonth2015 OCGS 3-83 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAI (cont)iS a s (

o n ti i.-. " .-_i 'i. ," : : i~i..=.III

.... L , ..

i.. .. .... ... .... .... ... .. i .. '... .. ... ... .... ..'.. .": " .i~ i....Basis (cont): " 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." 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 MS1.bBasis Reference(s):

1. NEI 99-01 Rev 6, SA1 2. UFSAR Section 8.2, Offsite Power System 2. OCNGS Drawing BR 3000 3. ABN-36, Loss of Off-Site Power 4. ABN-37, Station Blackout 5. ABN-60, Grid Emergency xxxlVlonth2015 OCGS 3-84 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU1.Initiating Condition:.., Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Ap licabil.ity:

1,2 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Loss of ALL offsite AC power capability to 4160V Buses 1C and 1 D for >15 minutes.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 potentialreduction 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.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 MA1.Basis Reference(s):

--..1. NEI 99-01 Rev 6, SU1 2. UFSAR Section 8.2, Offsite Power System 3. OCNGS Drawing BR 3000 4. ABN-36, Loss of Off-Site Power 5. ABN-60, Grid Emergency xxxlVIonth2015 OCGS 3-85 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 Initiating Condition:

Loss of all AC and Vital DC power sources for 15 minutes or longer.;OperatingMod ,eAppli~ca bili .ity:;, , .." ..1,2... ..... .~~~~~~~~~~~~~

... .. ....., , , ,: ........... , , ,. ....Emergency Act!o Lee1EAL):'

Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of ALL offsite AC power to 4160V BusesiC and 1iD.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D.AND 3. Voltage is < 115 VDC on 125 VDC battery busses B and C.AND 4. ALL AC and Vital DC power sources have been lost for > 15 minutes.,b a s is : ..... ... ..... ... ... .... ......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 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 all EAL conditions are met.xxxMonth2015 OCGS 3-86 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)Basis Reference(s):

1. NEI 99-01 Rev 6, SG8 2. UFSAR Section 8.3.2, DC Power Systems 3. UFSAR Section 8.2, Offsite Power System 4. OCNGS Drawing BR 3000 5. ABN-36, Loss of Off-Site Power 6. ABN-37, Station Blackout 7. ABN-60, Grid Emergency 8. ABN-54, Loss of DC Distribution Center B 9. ABN-55, Loss of DC Distribution Center C xxxIVlonth2015 OCGS 3-87 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2 Initiating Condition:

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

........ .......1,2 Emergency Action Level (EAL): ...Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Voltage is < 115 VDC on 125 VDC battery busses B and C for > 15 minutes.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 RG1, FG1 or MG2.R eferenc (-)-.`.__'

' ..... .....". ........_ :. ... ;- --" ... "' :, .... .' '.. ..' .. ..".,I....

.. ... .... ;::. .. .T --.t Basis Reference(s):.-

1. NEI 99-01 Rev 6, SS8 2. OCNGS Drawing BR 3000 3. ABN-54, Loss of DC Distribution Center B 4. ABN-55, Loss of DC Distribution Center C xxxMonth2015 OCGS 3-88 EP-AA-1010 Rev XX Ovstar Crook Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 mInitia"t-in-g Condition:

Inability to shutdown the reactor causing a challenge to RPV Water Level or RCS heat removal.Operating Mode Applicability:~

1 Em ergency, Action evel AL): .... .......................

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND 2. ALL manual I ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 2%.AND 3. EITHER of the following conditions exist: " RPV water level cannot be restored and maintained

> -20 inches TAF.OR* Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded......... .... .. I. ...... ..... ....., ..: ....... .: ......-..*. ... ........ ..Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection, are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.

This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.EAL #3 is considered to be exceeded when, as specified in the site-specific EOPs, RPV water level cannot be restored and maintained above the specified level.Escalation of the emergency classification level would be via IC RG1 or FG1.xxxlVlonth2015 OCGS 3-89 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)Basis Reference(s):

1. NEI 99-01 Rev 6, SS5 2. EMG-3200.01 B, RPV Control -with ATWS 3. EMG-3200.02, Primary Containment Control 4. 2000-BAS-3200.02, EOP User's Guide 5. 2000-GLN-3200.01, Plant Specific Technical Guideline 6. EMG-3200.01A, RPV Control -no ATWS xxxlVlonth2015 OCGS 3-90 EP-AA-1 010 Rev XX (I *+or ('roank tZoarnfrinn -qh=itin Annov F~n ~II~ ,**5 .4U* *Pvi:: n=rn MN i ~lnr L Z3 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 initiating Condition:

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

1!E m e rg e n C ~y ..... L e. e .i i ... ...-.-.......

..... .-.. .i i. ; ............

....... .:. ......... .... ... .*.. ..... ..... ... ..... ... ... ... ...... ... .........

... .:. .:... ................

..* ,. ...... ..........

.* ..... ...- ----... ..Emergency Action Level (MAL): 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. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 2%.Bfa-s is: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor consoles to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor consoles since this event entails a significant failure of the RPS.A manual action at the reactor consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor consoles (e.g., locally opening breakers).

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.xxxMonth2015 OCGS 3-91 EP-AA-1010 Rev XX tl afa r*an I'an rni A*fn ** *** tt* 5**N,,rltnr Z 23 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 (cont)Basis (cont):...".

..The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MS3. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC MS3 or FS1, an Alert declaration is appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.B a Is ~ f r n -~ ... .... .. ... .... .... 7 :. .... ...... ..... ........ ... ..: ...... .... .... .... ....... ....... .... .-- : :; i: .... ....basis Reference(s):

-1. NEI 99-01 Rev 6, SA5 2. EMG-3200.01A, RPV Control -no ATVVS 3. EMG-3200.01 B, RPV Control -with ATWS 4. 2000-BAS-3200.02, EOP User's Guide 5. 2000-GLN-3200.01, Plant Specific Technical Guideline xxxIVIonth2015 OCGS 3-92 EP-AA-1010 Rev XX (I afar (rookf'Zonar2fin -Qfnfinn nnny l::YlInn Mkil~r-l Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 Initiatin:

Condition:

Automatic or manual scram fails to shutdown the reactor.Op.eratin.g.o...

A.pplicability:........................

.1 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. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND b. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.OR 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power> 2%.AND b. EITHER of the following:

1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.OR 2. Subsequent automatic scram I ARI is successful in shutting down the reactor.This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, xxxMonth2015 OCGS 3-93 EP-AA-1010 Rev XX rN afgr fradmIe f'_nngarnfin

_Qfnfitn Anndav I=valrnn H~i ,ela r I~ ksI ~n~~in~~i,~

nF~InM~42 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)Basis:(cont)::

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 take manual action at another location(s) on the reactor consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies:

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plantconditions, etc. If subsequent operator manual actions taken at the reactor consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MA3. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC MA3 or FAI, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied." If the signal generated as a result of plant work causes a plant transient that creates a real condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

If the signal generated as a result of plant work does not cause a plant transient but should have generated an RPS scram signal and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

xxxMonth2015 OCGS 3-94 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)Basis Reference(s):

1. NEI 99-01 Rev 6, SU5 2. EMG-3200.01A, RPV Control -no ATWS 3. EMG-3200.01 B, RPV Control -with ATWS 4. 2000-BAS-3200.02, EOP User's Guide 5. 2000-GLN-3200.01, Plant Specific Technical Guideline xxxlVlonth2015 OCGS 3-95 EP-AA-1 010 Rev XX Ovater Creek Generatina Station Annex Exelon Nuclear Oyster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 Initiating Condition:.-.

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.operating Mode Applicability:

............

.... .... .......1,2 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for > 15 minutes.Table M1 Control Room Parameters

Reactor Power" RPV Water Level* RPV Pressure* Drywell Pressure* Torus Water Level* Torus Water Temperature AND 2. ANY Table M2 transient in progress.Table M2 Significant Transients" Turbine Trip* Reactor Scram* ECCS Actuation* Thermal power change > 25%* Thermal Power oscillations

> 10%xxxIVlonth2015 OCGS 3-96 EP-AA-1010 Rev XX Ovster Creek Generatinci Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 (cont)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 any 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.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency~classification level would be via ICs FS1 or IC RS1.1. NEI 99-01 Rev 6, SA2 xxxMonth2015 OCGS 3-97 EP-AA-1010 Rev XX Ovstar Creek Generatina Station Annex Exelon Nuclear Oyster..C.eek..G.n.r.t.n.

Station..Anne..E....on.Nu.lear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4 Initiatin-gCondition:-

'-, UNPLANNED loss of Control Room indications for 15 minutes or longer..Oe a i g M d ~ p ~ i a ii y : ........ .... .........

....... ....: ..... ... ..... .... ... ' ........... Operating .Mode Applicability:

1,2.EmergencyAction Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.UNPLANNED event results in the inability to monitor ANY Table M1 parameters from within the Control Room for > 15 minutes.Table M1 Control Room Parameters

Reactor Power* RPV Water Level* RPV Pressure" Drywell Pressure* Torus Water Level" Torus Water Temperature i~ ~s s .. i: ; : ....... .... ; ..;. ;... ... ... ; :.... ,_ {. , ::i !;... ... .BIasis-:',,-,-.

UNPLANNED:

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

The cause of the parameter change or event may be known or unknown.SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for any 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.xxxMonth2015 OCGS 3-98 EP-AA-1010 Rev XX t-% afnr Prdanif flnnnrnfin -Q+ofirn nndav I::Y-lf~n NKl,,alnr z~ ~ ~~~iNet~~rnA

~ v~,nMuI Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4 (cont)Basis. (cont): ..An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event would be reported if it significantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC MA4.Basis Reference(s):

... ... ... ... ..... ... .... ." ......1. NEI 99-01 Rev 6, SU2 xxxMonth2015 OCGS 3-99 EP-AA-1010 Rev XX (I atop f'rimak (' Zongarnfin -qfn+itn Anndav I=valrn Kim inilr w V -. 5* * *.* .** si L~4L*.

fl* ** *fl t

U * * ~ -~*Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.Operating Mode Appi ca ility:......

1,2.. .. ~ ~ ~ ~ ...... .......:; ...;:./ .." .." ....Emergency Action Level (EAL-):.........

Note:* If it is determined that the conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.1. The occurrence of ANY of the following hazardous events: " Seismic event (earthquake)" Internal or external flooding event* High winds or tornado strike" FIRE" EXPLOSION" Other events with similar hazard characteristics as determined by the Shift Manager AND 2. EITHER of the following:

a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.OR b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical Specifications for the current operating mode.G. A SeiSMic event required a manual reactor Sc~ram per ABN 38-, Station Seismic Eivent.Bas is: 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 large quantities of smoke and heat are observed.xxxMonth2015 OCGS 3-100 EP-AA-1010 Rev XX I=xelon Nuclear flv-tar C~rook Genersatinn Station Annre ExeIon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 (cont)Basis (cant)ý: 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 energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such 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 ECCS. 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, 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 equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in 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 #2.b 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 operation or readily apparent through indications alone, 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 RS1.If the EAL conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.p as..is.,Reference(s):

1. NEI 99-01, Rev 6 SA9 2. ABN-38 xxxMonth2015 OCGS 3-101 EP-AA-1010 Rev XX r) a+nr f'rAmAnk t'-dndnr!m+in -Q+!m+inn Annnv I='Yml^ n MI ,il,'mr Aua+~r (~ r~~~Ii f~~n~r~4i nil ~4~+ivhn An n~v kim ~ *OL~.* * ~ * ** SL** *1 tt4L*.

rt* ** *~a fl~ * * * * ~ .A *t*Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 InitiatifgýC, Condition:

RCS leakage for 15 minutes or longer.Opperating

_Mode Applicabiiy 1,2;Emergejncy Action Level (PAL):., Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for> 15 minutes.OR 2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.OR 3. Leakage from the RCS to a location outside the Drywell >25 gpm for > 15 minutes.Bais-: --UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

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 or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.

Lesser values typically require time-consuming xxxlVlonth2015 OCGS 3-102 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)Basis (cnnt): 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 any relief valve does not warrant an emergency classification.

A stuck-open Electromatic Relief Valve (EMRV) or EMRV leakage is not considered either identified or unidentified leakage by Technical Specification and, therefore, is not applicable to this EAL.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 ICs of Recognition Category R or F.Basis Reference s):..:e .. .... --1 11 ý .1.... ..' .i.. ... -1. NEI 99-01 Rev 6, SU4 2. Technical Specifications 3.3.D, Reactor Coolant System Leakage xxxMonth2015 OCGS 3-103 EP-AA-1010 Rev XX Oyster Creek Generatingc Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Iinitiatijng Condition:

Loss of all On-site or Off-site communications capabilities.

Operating Mode Applicability:

1,2 ELmergency Action Level (EAL): 1. Loss of all Table M3 Onsite communications capability affecting the ability to perform routine operations.

OR 2. Loss of all Table M3 Offsite communication capability affecting the ability to perform offsite notifications.

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

Table M3 Communications Capability System Onsite Offsite NRC Plant Paging System X Station Radio X Conventional Telephone lines X X X Satellite Phone system X X X NARS X HPN X X ENS X X.Basis: This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).xxxlVlonth2015 OCGS 3-104 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)Basis'(cant):

EAL #1 Basis Addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are listed in procedure EP-MA-1 14-100-F-03, State/Local Notification Form.EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Basis Reference(s):.

1. NEI 99-01 Rev 6, SU6 2. UFSAR Section 9.5.2, Communication Systems 3. EP-AA-124-1001, Facilities Inventories and Equipment Tests xxxlVlonth2015 OCGS 3-105 EP-AA-1 010 Rev XX Ovster C~reek (GAneratinn Station Annex FYAlnn Nulair~I Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA1 Initating Con.dition:

Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer.Operain hg:ModIVei Applicability:

3,4, D Emergency (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of all offsite AC power to4160V Buses 1C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D..AND 3. Failure to restore power to at least one 4160V bus (1C or 1D) 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.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via IC CS6 or RS1.xxxMonth2015 OCGS 3-106 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CAI (cont)Basis Reference(s).:...

1. NEI 99-01 Rev 6, CA2 2. ABN-36, Loss of Off-Site Power 3. ABN-37, Station Blackout 4. ABN-60, Grid Emergency 5. OCNGS Drawing BR 3000 6. UFSAR Section 8.2, Offsite Power System xxxlVlonth2015 OCGS 3-107 EP-AA-1010 Rev XX flv~tc~r (~ri~k (~nAr~tinn

~t~mtinn Anncw Fxelnn r Ovster Creek Generatinn Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU1 tiating Condition.i.

.... ........ .................

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

.. ..... ........ ....... ........ .3,4, D Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. AC power capability to 4160V Buses 1C and 1 D reduced to only one of the following power sources for > 15 minutes.* Startup Transformer SA" Startup Transformer SB* EDG-1 Emergency Diesel Generator* EDG-2 Emergency Diesel Generator AND 2. ANY additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.B a s is : ........ ... ... .......SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.xxxMonth2015 OCGS 3-108 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CUI (cont)InitiatingCondition:

An "AC power source" is a source recognized in ABNs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all 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." 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.The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA1.Basis Reference(s):

1. NEI 99-01 Rev 6 CU2 2. ABN-36, Loss of Off-Site Power 3. ABN-37, Station Blackout 4. ABN-60, Grid Emergency 5. OCNGS Drawing BR 3000 6. UFSAR Section 8.2, Offsite Power System xxxlVlonth2015 OCGS 3-109 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Ostr1 Crek.Gne.aina Sttin. nn.E..o.Ncla Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA2...... .... ... .. ....: , ..... .-C A ..Initiating Condition, Hazardous event affecting SAFETY SYSTEM required for the current operating mode.Operating MoeApica~bility'::

-g _ _ _ _ e .... ... -----... ........, j. ......, , : .. , : , : , .... ... .... .., = ...... ... .... .. ........: .. .....3,4 Emergency Action Level(EAL):

.Note:* If it is determined that the conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.1. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)" Internal or external flooding event* High winds or tornado strike" FIRE" EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager AND 2. EITHER of the following:

a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.OR b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical Specifications for the current operating mode.OR G. A seismic event required a manu-l reactor .cr.m pcr AN 38 Station Scismic Event B a S : : : ..... : : .. ....... .. .. .... ..... : .... ..... .. .": ... ........... .... ....... ... .......Bas~is: 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 xxxMonth2015 OCGS 3-110 EP-AA-1 010 Rev XX Ovater Creek Generatina Station Annex Exellon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA2 (cont)basis- (cont):...........................

.flame is preferred but is NOT required if large 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 energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such 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 ECCS. These are typically systems classified as safety-related.

VISIBLE DAMAGE: Damage to acomponent 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, 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 equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in 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 #2.b 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 operation or readily apparent through indications alone, or 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 CS6 or RS1.If the EAL conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.B a s is R e fe r n c e ~ s : " " " ...... ...... ..........

' ........ " ... ..... ...... ..... ........ ....................." ...................

Basis Reference(s):

6, 1. NEI 99-01 Rev 6, CA6 xxxMonth2015 OCGS 3-111 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS

2. ABN-38, Station Seismic Event 3. FSAR Update Section 3. 7 (Seismic)4. FSAR Update Section 3.3.1 (High winds)5. ABN-31, High Winds 6. ABN-32, Abnormal Intake Level 7. ABN-29, Plant Fires xxxlVlonth2015 OCGS 3-112 EP-AA-1 010 Rev XX r) afar rrank rZannrntin Atnfinn Annoy Fx#_lnn NuleIasr Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis.RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:

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

3,4 Emeirgncy Actio6Lev (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Voltage is < 115 VDC on required 125 VDC battery busses B and C for> 15 minutes.B a sis : ...... ......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 operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.

For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level would be via IC CA6 or CA5, or an IC in Recognition Category R.Baisis.ReferienCce(s):

1. NEI 99-01 Rev 6, CU4 xxxMonth2015 OCGS 3-113 EP-AA-1 010 Rev XX Fxnlnn N.P.Innr vsVtVr CIreek (GAneratinn Station *nncex Ex,.oin NuiclIeaVVr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS

2. UFSAR Section 8.3.2, DC Power Systems 3. OCNGS Drawing BR 3000 4. ABN-54, DC Bus B and Panel/MOC Failures 5. ABN-55, DC Bus C and Panel/MCC Failures xxxMonth2015 OCGS 3-114 EP-AA-1010 Rev XX f) a+nr (rnnk I'Zandarnfin -q+!%+irn Annow FIvcnn Mlm ,alnr w ~OL..I -~ * ** 4L** si ~S**

fl* ** *fl -... ----.Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU4 itatian g Condition:

Loss of all onsite or offsite communications capabilities.

operating Mode Applicability:

3,4, D Emergency Action Level (EAL): 1. Loss of all Table C1 Onsite communications capability affecting the ability to perform routine operations.

OR.2. Loss of all Table C1 Offsite communication capability affecting the ability to perform offsite notifications.

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

Table Cl -Communications Capability System Onsite Offsite NRC Plant Paging System X Station Radio X Conventional Telephone lines X X X Satellite Phone System X X X NARS X HPN X X ENS X X Basis: This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).EAL #1 Basis Addresses a total loss of the communications methods used in support of routine plant operations.

xxxlVlonth2015 OCGS 3-115 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 (cont)Basis (cont)0 EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are listed in procedure EP-MA-1 14-100-F-03, State/Local Notification Form.EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Basis Reference(s):_

1. NEI 99-01 Rev 6, CU5 2. UFSAR Section 9.5.2, Communication Systems 3. EP-AA-124-1001, Facilities Inventories and Equipment Tests xxxIVIonth2015 OCGS 3-116 EP-AA-1 010 Rev XX F:yinlnn Miion~lr COvster Cree~k Generatinn Station AnnexFA~nNirI~

Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA5' :". ". "" ..... ... ...I. ..... ....; " .. ..... .Initiating Co.ndition.......:.

Inability to maintain the plant in cold shutdown.Operating 3,4 EmergencyAction Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.

1. UNPLANNED rise in RCS temperature

> 2121F due to loss of de.femoval-for

> Table C2 duration.Udy mlfddt Table C2 RCS Heat-up Duration Thresholds RCS Containment Closure Heat-up Status Status Duration Intact Not Applicable 60 minutes*Established 20 minutes*Not Intact Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, then EAL #1 is not applicable.

OR 2. UNPLANNED RPV pressure rise > 10 psig as a result of temperature rise-due-4e loss of decay hecat Frcmoval.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.xxxlVlonth2015 OCGS 3-117 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA5 (cont)Basis (cont): .. ....CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise.The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature rise without a substantialdegradation in plant safety.Finally, in the case where there is a rise in RCS temperature, the RCS is not intact, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).

This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.EAL #2 provides a pressure-based indication of RCS heat-up.Escalation of the emergency classification level would be via IC CS6 or RS1.B asis R efe re nce(s),..

..... ......... ................

............

1. NEI 99-01 Rev 6, CA3 2. Technical Specifications 1.7 3. OU-AA-1 03, Shutdown Safety Management Program xxxMonth2015 OCGS 3-118 EP-AA-1 010 Rev XX (I afar rrn C* nrni qfqi* **An *

C.Pyninn N,,r~lgr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU5 i-nitiating Condition:

UNPLANNED rise in RCS temperature OperatingMode Applicability:

---3,4'Emergency Action Level (E AL): " Note: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.

1. UNPLANNED rise in RCS temperature

> 212°F due to loss of decay heat Femoval.OR 2. Loss of the following for >15 minutes." ALL RCS temperature indications AND* ALL RPV water level indications.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.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA5.xxxMonth2015 OCGS 3-119 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU5 (cont)..... ..... .. : .. .; , :... ... .... ., ..... ... L ; ...... ..

., .., ..." , .......Basis (cont): RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

EAL #1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.

A loss of forced decay heat removal at reduced inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown.EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA6 based on an inventory loss or IC CA5 based on exceeding plant configuration-specific time criteria.Basis' Reference(s):

a .. ...... ......... ....... ... * ..* .s ) ..... ... .... .... ...: .... ... ....... ..........

.. _... .....-.1. NEI 99-01 Rev 6, CU3 2. Technical Specifications 1.7 3. 2000-GLN-3200.01, Plant Specific Technical Guideline 4. 205.94.0, RPV Floodup Using Core Spray 5. 205.95.0, Reactor Flood-up / Drain-down xxxMonth2015 OCGS 3-120 EP-AA-1010 Rev XX I=xelon Nuclear C(--,-tr Creek (GAneratinn Station Annex Exelon Nucle~ar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 initiating Condition:, Loss of RPV inventory affecting fuel clad integrity with containment challenged..Operating Mode. Applicabilitf.y:

3,4 Emeigency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. a. RPV water level < 0 inches TAF for > 30 minutes.AND b. Any Containment Challenge Indication (Table C4)OR 2. a. RPV water level unknown cannot be monitored for > 30 minutes.AND b. Core uncovery is indicated by ANY of the following:

Table C3 indications of a sufficient magnitude to indicate core uncovery.OR* Refuel Floor Area Radiation Monitor C-1 0, North Wall, reading>3 R/hr.AND c. ANY Containment Challenge Indication (Table C4)Table C3 Indications of RCS Leakage" UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

xxxMonth2015 OCGS 3-121 EP-AA-1010 Rev XX Fxnlnn Ovster ('raak Gana~rnting Station Annexv EaeInn Nucler~I Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)Emergency Action Level (EAL) (cant): Table C4 Containment Challenge Indications

Primary Containment Hydrogen Concentration

> 6% and Oxygen Concentration

> 5%* UNPLANNED rise in containment pressure* CONTAINMENT CLOSURE not established*, Any Secondary Containment radiation monitor > EMG-3200.11 Maximum Safe (Table 12).* 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.B a s i : .... 7 ..... .........

........ , ..... ... ..... ... ..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.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.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary for BWR) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.

This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA Protective Action Guidelines (PAG) exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.xxxMonth2015 OCGS 3-122 EP-AA-1 010 Rev XX Nuclear rl cfkr C~rook r~onerzpting S~tation Annex Exelon Nuclea~r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)'bBasis (cont):............-.

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.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 challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.

If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access.During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

EAL #1 Basis The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).

It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.

If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or Torus levels. Sump and/or Torus level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

xxxMonth2015 OCGS 3-123 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN REFUELING SYSTEM MALFUNCTIONS CG6 (cont)Bas!s Reference(s):

1. NEI 99-01 Rev 6, CG1 2. EMG-3200.01A, RPV Control -No ATWS 3. Technical Specifications

5.2 Basis

4. 2000-GLN-3200.03, Plant Specific Technical Guidelines for Severe Accident5Guidelines

5. EMG-3200.1 1, Secondary Containment Control 6. FSAR Update 6.2.1.1.3 7. EMG-3200.02, Primary Containment Control 8. EP-AEL-0501, Estimation of Radiation Monitor Readings Indicating Core Uncovery xxxMonth2015 OCGS 3-124 EP-AA-1 010 Rev XX

() afar rrook rýanarntin Af2finn Annoy (h~cor C~~Ic(~n~~tin

~t~inn nn~vFxalnn Nucleiaar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 Initiating condition-:

Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

.-...3,4 Emre r g-e-nc6y-A-c tiobn -L e've61 (-E-AL0): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. With CONTAINMENT CLOSURE not established, RPV water level < 56 inches TAF.OR 2. With CONTAINMENT CLOSURE established, RPV water level < 0 inches TAF.OR 3. a. RPV water level cannot be monitoredunknown for > 30 minutes AND b. Core uncovery is indicated by ANY of the following:

Table C3 indications of a sufficient magnitude to indicate core uncovery.OR* Refuel Floor Area Radiation Monitor C-10, North Wall, reading >3 R/hr.Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

xxxMonth2015 OCGS 3-125 EP-AA-1010 Rev XX flv~ti~r Cr~~k (~n~r2tinn

~t2tinn Ann~v r C) ctor rtraak rZanarnting Stntinn nnax Explon Nuclear.L Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 (cont)* ~~ ~ ~ ~ ~ ~ ... .........

.,*.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.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary for BWR) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions.

The difference in the specified RCS/reactor vessel levels of EALs #1 and #2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

In EAL #3.a the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).

It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.

If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or Torus levels. Sump and/or Torus level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG6 or RGI.xxxMonth2015 OCGS 3-126 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 (cont)Basis Reference(s):

1. NEI 99-01 Rev 6, CS1 2. Technical Specification Table 3.1.1 3. EMG-3200.01A, RPV Control -No AT-WS 4. 2000-GLN-3200.03, Plant Specific Technical Guidelines for EOPs 5. EP-EAL-0501, Estimation of Radiation Monitor Readings Indicating Core Uncovery xxxMonth2015 OCGS 3-127 EP-AA-1 010 Rev XX C) thir C~reak Ge0nerating Station Annaev Fveln Nui~clear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:

Loss of RPV inventory.

Ope~ra~ting.gMode Applicability:

3,4 Emergen cy Acton Level (EAL): Note: 0 The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. a. Loss of RPV inventory as indicated by level < 86 inches TAF.OR 2. a. RPV water level cannot be monitoredunknown, for> 16 minutes.AND b. Loss of RPV inventory per Table C3 indications.

Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise*" UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

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 conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).

This condition represents a potential substantial reduction in the level of plant safety.EAL #1 Basis A lowering of water level below 86 inches indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery.xxxlVlonth2015 OCGS 3-128 EP-AA-1 010 Rev XX flvs~ter Creek (GAneratinn Sta~tion Annex Exelon Nucleair Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA6 (cont)Basis (cant): Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). A rise ,in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA5.EAL #2 Basis The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation.

If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or Torus levels. Sump and/or Torus level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS6 If the RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS6.Basis Reference(s):

1. NEI 99-01 Rev 6, CA1 2. Technical Specification Table 3.1.1 3. 2000-BAS-3200.02 EOP User's Guide 4. 2000-GLN-3200.01 Plant Specific Technical Guideline 5. 205.94.0 RPV Floodup Using Core Spray 6. 205.95.0 Reactor Flood-up / Drain-down

7. FSAR Figure 7.6-3 xxxlVlonth2015 OCGS 3-129 EP-AA-1010 Rev XX flu~t~r Cr~~k (~n~r2tinn

~hitinn Ann~v Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU6 initiating Condition.:..

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

3,4 Emergency Action Level (EAL) 'Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. UNPLANNED loss of reactor coolant results in the inability to restore and maintain RPV water level to above the procedurally established lower limit for> 15 minutes.OR 2. a. RPV water level cannot be monitoredufknown AND b. Loss of RPV inventory per Table C3 indications.

Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise*" Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

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 the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV water level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.xxxIVlonth2015 OCGS 3-130 EP-AA-1010 Rev XX rl afar 1'rack rZanarnfin Atntinn Annoy A~,cicrPl vn~r.ale (~lnlabrtin

¥t~t4*II Anl~ lW In Nu.cleaI r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU6 (cont)Basis" (cant): The procedurally established lower limit is not an operational band established above the procedural limit to allow for operator action prior to exceeding the procedural limit, but it is the procedurally established lower limit.Refueling evolutions that lower RCS water inventory are carefully planned and controlled.

An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.EAL #1 Basis Recognizes that the minimum required RPV water level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented.

This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water ,level. This criterion excludes transient conditions causing a brief lowering of water level.EAL #2 Basis Addresses a condition where all means to determine RPV water level have been lost.In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or Torus levels. Sump and/or Torus level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA6 or CA5.Basis Refer'ence(s):

.........

...1. NEI 99-01, Rev. 6 CU1 2. Technical Specifications Section 3.3.D 3. 205.94.0, RPV Floodup Using Core Spray 4. 205.95.0, Reactor Flood-up / Drain-down

5. FSAR Figure 7.6-3 xxxlVIonth2015 OCGS 3-131 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster.Cree GeeainaSaonAexElnNuer Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 Initiating Condition:

HOSTILE ACTION resulting in loss of physical control of the facility.Operating Mode Applicability:

.... ... .....1,2,3,4, D Emergency Action Lev-el (EAL):.1. A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.AND 2. a. ANY Table H1 safety function cannot be controlled or maintained.

OR b. Damage to spent fuel has occurred or is IMMINENT Table H1 Safety Functions* Reactivity Control (ability to shut down the reactor and keep it shutdown)0 RPV Water Level (ability to cool the core)0 RCS Heat Removal (ability to maintain heat sink)B as is :. " .... ...... ....HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.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.xxxMonth2015 OCGS 3-132 EP-AA-1010 Rev XX fl afnr Prniak IZinnarnfin A+nfinn Annoy IPYalnn NI Wy l5 *oL m * *5

ll~ .aal* t t. Lc~e(a~~inn LlVlm l Amnmav Fv n M* ** w Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 (cont)Basis (cont):'HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.

It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].Basis Reference(s):

1. NEI 99-01, Rev. 6 HG1 5. Station Security Plan -Appendix C xxxMonth2015 OCGS 3-133 EP-AA-1010 Rev XX n a+,mr 1rdnxlr flnndmrnfin Q+n+itn Annekv F:::volrn M, rl,-nr I~ r~I Vkn~~i r 44rnn v arnNi62 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HSl!In itiating C~ nditi~ n-i~i ..i. ... .... .. i .. .....~ ....

.. ~h .i -. i. ..i ......Initiating Condition-:'

HOSTILE ACTION within the PROTECTED AREA.Operating ModeApplicability:

.1, 2,3,4, D'Emergency Action Level (EAL): A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.Basis': HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].xxxMonth2015 OCGS 3-134 EP-AA-1010 Rev XX rN afinr (rganlr IZannrnfin -Qfnfie-.n Anngav I:valnn I~nl,, Z',e~ (:raI 2~~i n~~rnA ayFan irI Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HSI (cont)..... ... ,: ... Basis (cont):.As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The Site Area Emergency declaration will mobilize ORO resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAl. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR§ 50.72.Escalation of the emergency classification level would be via IC HG1.Basis Refere.nceS):

..... .. .--.1. NEI 99-01 Rev 6, HS1 3. Station Security Plan -Appendix C xxxMonth2015 OCGS 3-135 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster...

Creek ...e.e.at.na Statin.Annx.E..on.N.lea Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 Initiating Condition:

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.Operatinga Mode.Applicabiity:

1,2,3,4, D Action Level (EAL): 1. A validated notification from NRC of an aircraft attack threat < 30 minutes from the site.OR 2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE:.

A person(s) held as leverage against the station to ensure that demands will be met by the station.PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.OWNER CONTROLLED AREA (OCA): The property associated with the station and owned by the company. Access is normally limited to persons entering for official business.PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.xxxMonth2015 OCGS 3-136 EP-AA-1010 Rev XX t'l afar ('rank (Zanarnfin

_qfnfirn Annew IPYalnn Chc4~r ('~raai, t~nar~tinn

~t~tmnn Annchv Fv~Inn Miir~Ic~zar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAI (cont)Basis (cont): Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE.Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL #1 Basis Addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness.

This EAL is met when the threat-related information has been validated in accordance with ABN-41, Security Event.EAL #2 Basis Applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.

The status and size of the plane may be provided by NORAD through the NRC.In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.Escalation of the emergency classification level would be via IC HS1.xxxMonth2015 OCGS 3-137 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)Basis Refere-nce(s)-`:, 1. NEI 99-01 Rev 6, HA1 2. Station Security Plan -Appendix C 3. ABN-41, Security Event xxxMonth2015 OCGS 3-138 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 in itiati ng, i iC(ondition:,:..

..Confirmed SECURITY CONDITION or threat.Op-er ating_ Moide.Appllicability:

1,2,3,4, D EmergencyAction Level (EAL): 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. 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.a s ~s : , ..........

..:.....`..

..: ....:: ..M : ..... ..... ... £ ... ........: ........ .. .. ... .. ...:b a s'is :..... .............

SECURITY CONDITION:

Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION 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.

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety.Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.xxxMonth2015 OCGS 3-139 EP-AA-1010 Rev XX rN afgar PrdmekL- ('_dmnnr!2+in Q+n+ifn Anniav I=-.lx1rn NKlrelatr (h,~+~r I~ rn~I, (~anar,~4in,, Q*,,+i,%n A nn~v ~vaIr~n Mi.eI~2r-r0L. .n .s. ~I.

s~a.j.. n. I. .a n .~.. ..-- .~...Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU! (cont).Basis (cont): Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs.Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].EAL #1 Basis Addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with SY-AA-1 01-132.EAL #2 Basis Addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.

The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with ABN-41, Security Event.EAL #3 Basis References Security Force because these are the individuals trained to confirm that a security event is occurring or has occurred.

Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

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

...........

1. NEI 99-01 Rev 6, HUl 2. Station Security Plan -Appendix C 3. ABN-41, Security Event 4. SY-AA-101-132, Security Assessment and Response to Unusual Activities xxxMonth2015 OCGS 3-140 EP-AA-1010 Rev XX r% afmr (radak (%dandarnfin -Qfnfinn Annov R:v-lInn PlJ"Al-r z~@~ :' ~gr4in~2 ti~ n v*- U**Ut U** *4L**S **f l*Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating.pcondition:

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

................

...1,2,3,4, D Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per ABN-30 Control Room Evacuation.

AND 2. Control of ANY Table H1 safety function is not reestablished in < 15 minutes.Table H1 Safety Functions* Reactivity Control (ability to shut down the reactor and keep it shutdown)" RPV Water Level (ability to cool the core)* RCS Heat Removal (ability to maintain heat sink).Basis: The time period to establish control of the plant starts when either: a. Control of the plant is no longer maintained in the Main Control Room OR b. The last Operator has left the Main Control Room.This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plan control to alternate locations is a precursor to a challenge to any fission product barriers within a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment.

The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Escalation of the emergency classification level would be via IC FG1 or CG6.xxxMonth2015 OCGS 3-141 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)Reference(s):,....

1. NEI 99-01, Rev 6 HS6 2. ABN-30 Control Room Evacuation

3. Procedure 346 Operation of the Remote and Local Shutdown Panels xxxMonth2015 OCGS 3-142 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2 Initiating, Condition:

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

Operating Mode APplicability:'ý:,..

1,2,3,4, D Emergency Action Level (EAL):..A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per ABN-30 Control Room Evacuation.

Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.

The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel.

Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Escalation of the emergency classification level would be via IC HS2.Ba'sis". Reernce(s):.

1. NEI 99-01, Rev 6 HA6 2. ABN-30 Control Room Evacuation

3. Procedure 346 Operation of the Remote and Local Shutdown Panels xxxlVlonth2015 OCGS 3-143 EP-AA-1010 Rev XX Ovster Creek Generating Station Annex Exelon Nuclear OytrCee eeainqSainAnxEeo ula TabIA OC~S 3-2 OCGS EAL Technica Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 ,Initiating Condition:

FIRE potentially degrading the level of safety of the plant.Operating Mod'e -A'p'pilic'ability:

1,2,3,4, D Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* Escalation of the emergency classification level would be via IC CA2 or MA5 1 .A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of the following FIRE detection indications: " Report from the field (i.e., visual observation)

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

Field verification of a single fire alarm Table H2 Vital Areas 0 Reactor Building (when inerted the Drywell is exempt)0 4160V Switchgear Rooms (1C and 1D)& Control Room Complex (MOB, Upper and Lower Cable Spreading Rooms)* Main Transformer/Condensate Transfer Pad* Intake Structure* #1 EDG Vault 0 #2 EDGVault& EDG Fuel Oil Storage Tank OR 2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of a FIRE).AND b. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.OR 3 A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in< 60 minutes of the initial report, alarm or indication.

OR xxxMonth2015 OCGS 3-144 EP-AA-1010 Rev XX Exl~xnn NuirnlInr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY 4 A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

xxxMonth2015 OCGS 3-145 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)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 large quantities of smoke and heat are observed.PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.EAL #1 Basis The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed.

Similarly, the fire duration clock also starts at the time of receipt of the initial alarms, indication or report.EAL #2 Basis This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.xxxMonth2015 OCGS 3-146 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)Basis (qont): If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

EAL #3 Basis In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA.EAL #4 Basis If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded.

The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.

Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.ISFSI is not specifically addressed in EAL #3 and #4 since it is within the plant PROTECTED AREA and is therefore covered under EALs #3 and #4.Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

xxxMonth2015 OCGS 3-147 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)Basis (cont):.In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA2 or MA5.Basis Reference(i):

1. NEI 99-01, Rev 6 HU4 2. ABN-29, Plant Fires xxxMonth2015 OCGS 3-148 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4.... ... .... ... ...... ...........

..H U .4.Initiating Condition:.

Seismic event greater than OBE levels........li y : .... ... .. ....... .. .. .. ............ .. .... .... ........ .. .... .... ..... ... ... ..... ... ! :.pe~rating Mode Applicabil ity:-.....

1,2,3,4, D Emergency Action Level (EAL): Note:* Escalation of the emergency classification level would be via IC CA2 or MA5" For emergency classification if. EAL 2 is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in < 15 mins of the event.Seismic event as indicated by: 1. Control Room personnel feel an actual or potential seismic event.AND 2. ANY one of the following confirmed in < 15 mins of the event: " The earthquake resulted in Modified Mercalli Intensity (MMI) > VI and occurred< 3.5 miles of the plant." The earthquake was magnitude

> 6.0" The earthquake was magnitude

> 5.0 and occurred < 125 miles of the plant.Seeis4G--as indAcated6-by 1 Control Room pe..o..nl feel an actual or potential , eismic event.AND 2. The occurr~enco of -A seism~ic-event has resulted inaspurious ReactorF Sr-amFA Or required a Plant shutdo inacrdance with ABN 38, Station SeiSmicG Event.,B a s is : ........ .- ... ............

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)'. An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE)2 should have no significant impact on safety-related systems, structures and components; however, 1 An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the?ublic will remain functional.

An SSE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.

xxxMonth2015 OCGS 3-149 EP-AA-1 010 Rev XX Ovater Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections).

Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are in excess of 0.08g).EAL #2 and the accompanying note is included to ensure that a declaration does not result from felt vibrations caused by a non-seismic source (e.g., a dropped load). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., call to USGS, check internet source, etc.) however, the verification action must not preclude a timely emergency declaration.

This guidance recognizes that it may cause the site to declare an Unusual Event while another site, similarly affected but with readily available OBE indications in the Control Room, may not.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA2 or MA5.xxxMonth2015 OCGS 3-150 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)Basis Reference(s):

1. NEI 99-01, Rev 6 HU2 2. ABN-38, Station Seismic Event 3. FSAR Update Section 3. 7 (Seismic)xxxMonth2015 OCGS 3-151 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 Initiating Condition:

Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown..Oe a t n ..... ........ ....c a.b...i....i.

.. ...... ............... .x ... .... ... ... ...: :": ...... ..: !,Operating M ode Applicability:, .... ..... ...... ............

7-I2-,-3, 4,-9 Emergency Action Level (EAL): Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted.

1. Release of a toxic, corrosive, asphyxiant or flammable gas in a Table H3 area.Table H3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Reactor Building 51' elevation*

Modes 3 and 4*Arcpc required to shutdoWn coling AND 2. Entry into the room or area is prohibited or impeded Basis: .. ......This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal plant procedures.

This condition represents an actual or potential substantial degradation of the level of safety of the plant.Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR, therefore Table H3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.

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

xxxMonth2015 OCGS 3-152 EP-AA-1010 Rev XX Oyster Creek Generatinq Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)Basis (cont): ... ....... .This Table does not include the Control Room since adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.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 gaseous release preclude the ability to place shutdown cooling in service. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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 gaseous release).

For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing)." 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.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.

This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that generate smoke, that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

The Operating Mode Applicability of this EAL has been revised from All Modes to modes 3 and 4 due to the mode applicability of the areas of concern in Table H-3. In the xxxMonth2015 OCGS 3-153 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY future should the areas of concern in Table H-3 be revised then the Operating Mode Applicability of this EAL should be reevaluated.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICsas is... .Rfer n e ).. .. ... ...... : :: -'. iii:_.. .... .

basis Reference(s):

--1. NEI 99-01, Rev 6 HA5 2. ABN-33, Toxic or Flammable Gas Release 3. AR 660892, Station Halon and IDLH Evaluations

4. 29CFR1 910.134(b) and 29CFR1 910.134(d)(2)(iii) xxxMonth2015 OCGS 3-154 o EP-AA-1010 Rev XX rl a*O a ft rma

w* -n * ~

b** -qtt S 5

Annoy I=vjlnn I1i ,rlnr Z :a Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 iinitiating c o nditio n ~ ... .. .'.. ... ...i .i i ,l ... ..... .. .~ .. ............

.. i initiating Condition:

Hazardous Event Ope.rating..Mode Applicability::

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

Escalation of the emergency classification level would be via IC CA2 or MA5 1. Tornado strike within the PROTECTED AREA.OR 2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode.OR 3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).OR 4. A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.OR 5. Abnormal Intake Structure level, as indicated by EITHER: a. > 4.5 ft. MSL (> 4.25 psig on PI-533-1172 and PI-533-1173 or > 4.5 ft MSL on CR-423-11 pt 24 and pt 23).OR b. < -3.0 ft. MSL L< 0.95 psig on PI-533-1172 and PI-533-1173 or < -3.0 ft MSL on CR-423-11 pt 24 and pt 23).MSL = Mean Sea Level.B a s is : ..... ....... .... .... .... .. ........... .... ..PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.xxxMonth2015 OCGS 3-155 EP-AA-1 010 Rev XX C) a~car (iragde f(Znarnztinn -qtntinn AnnoyFvlnIJiI~r-,Ct*

'I *.

s-** *f F:Yalnn N"nionr Z Z3 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY 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.

'Basis (cont): This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1 Basis Addresses a tornado striking (touching down) within the Protected Area.EAL #2 Basis Addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.Classification is also required if the water level or related wetting causes an. automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. Manual isolation ot power to a SAFETY SYSTEM component a-. a result of leakage i, an evcnt of lesser imFpact and would be expected to cause small and locvaliZed-damnage. The consequnc of this type of event is adequately assessed and addressed in accordance with TeGhnial EAL #3 Basis Addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.EAL #4 Basis Addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.

Examples of such an event include site flooding caused by a hurricane, heavy rains, or high winds.This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.EAL #5 Basis: High Intake Structure level, is sufficiently high to require plant shutdown per ABN-32, Abnormal Intake Level.Low Intake Structure level indicates the possible loss of Radwaste Service Water pumps and is approaching levels, which may result in a loss of vital cooling equipment.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, H or C.xxxMonth2015 OCGS 3-156 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)Btasis Reference(s):

1. NEI 99-01, Rev 6 HU3 2. FSAR Update 3.3.1 (High winds)3. ABN-31, High Winds 4. ABN-32, Abnormal Intake Level xxxMonth2015 OCGS 3-157 EP-AA-1010 Rev XX t') atar (rrcank ('Znarnfinn -qtntinn AnnoyFaInJi~~~

~ S*4. 3 ~r Expinn Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG7 initiating Condition:..

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY.

Operating Mode Applicability:

.. ....= ....'. ... ....= .= ." 1,2,3,4, D.Emfergency Action Level (EAL): .. ........Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT 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 the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.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.HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.

Basis Reference(s):.

...1. NEI 99-01, Rev 6 HG7 xxxMonth2015 OCGS 3-158 EP-AA-1010 Rev XX t') afar ('raak rZontarnfin Atntinn nnay Pyl:YInn Nn n-lonr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Innnat g ond tno ; .: : .: -: .... ..... !/ .... ... ..... ... ... *S Initiatling' Co-nditio n: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY.

... ...... ... .. -...... ... .. ... ........ ... .: ..... .. .. ... .... ... ...Operating Mode Appllicability:;.

1,2,3,4, D Emer rgencyAction Level (EAL) : : ..........

....... ........ ..... ..Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major 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 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 expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.ýB a s is : .... ....... .... .............

.. ....HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.

Basis Reference(s):

1. NEI 99-01, Rev 6 HS7 xxxMontfi2015 OCGS 3-159 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Ilnitfiating Condition:..

Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT.Operating Mode Applicability:

...........

... ....1,2,3,4, D Eerge~cy.Ac~tion Level (EAL).:-Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels..Basis: HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE:

An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert.Basis Reference(s):

1. NEI 99-01, Rev 6 HA7 xxxMonth2015 OCGS 3-160 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU7.Initiating Condition:

'- ...Other conditions exist which in the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.Operating Mode Applicabillity:

.. ..... ... .....1,2,3,4,D=Emergency.

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

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Basis:A: .......This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENT..... .......... ..~~~~ ~ ~ ~~~~~~~~~

... ... .......... .....". .... .. ... .. ...:. ... ....., /. _ .o*. ..,Basis Reference(s):

1. NEI 99-01, Rev 6HU7 xxxMonth2015 OCGS 3-161 EP-AA-1010 Rev XX Pyalnn Ns inlanr () --far (rtrpk (Ganarzuing Rtzation Annex vIn ~~ca Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition

... .......... :.n t ~ t i g C n i i n ... ...... .. ...... ..... ....... -. ... ... .... i _ ..... ... ..... ..........

... .... .... ... ..... ... .. .. ...Damage to a loaded cask CONFINEMENT BOUNDARY.Operating Mode Applicability,:

1,2,3,4, D EmergencyAction:Level (EAL): .ii Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading:& > 800mr/hr (gamma + neutron) 3 feet from the HSM surface OR* > 200mr/hr (gamma + neutron) outside the HSM door on centerline of DSC OR* > 40mr/hr (gamma + neutron) end of shield wall exterior i.B a s is : .... .... ... .. ....... .. .. .. ... ......: ' ..........

.... ............ ... .... ....... * .. ....... ...... .. .... .. ..... ... ....Basis: CONFINEMENT BOUNDARY:

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

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) : A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The word cask, as used in this EAL, refers to the storage container in use at the site for dry storage of irradiated fuel. The issues of concern are the creation of a potential or actual release path to the environment, degradation of any fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions.

The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSIs are covered under ICs HU1 and HA1.xxxMonth2015 OCGS 3-162 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 (cont)Basis Reference(s):

1. NEI199-01, Rev 6 E-HU1 2. Oyster Creek Generating Station ISFSI 10CFR72.212 Evaluation Rev. 5 xxxlVlonth2015 OCGS 3-163 EP-AA-1010 Rev XX Oyster Creek Generatinq Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RG1 initiating Con'dition:

Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRem TEDE or 5000 mRem thyroid CDE.pr a t n M i p icb-i y i .... .. .. :::'i i l --oI. ...1 .-1 11 i -ý.:-_ -. .. .. .... ............

...... ...... :- ,." :" -...........

...... .......O.perating Mod~e Applicability:

1,2,3,4, D Emergepncy Action Level (EAL): ............

... ..Notes: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time." 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. 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. > 1000 mRem TEDE OR b. > 5000 mRem CDE Thyroid OR 3. Field survey results at or beyond the site boundary indicate EITHER: a. Gamma (closed window) dose rates >1000 mR/hr are expected to continue for > 60 minutes.OR b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.

xxxMonth2015 OCGS 3-33 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear OytrCee eeainaSainAnxEeo ula Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG1 (cont)Emergency ActionLeve (EAL) (cont):, Table RI Effluent Monitor Thresholds Effluent Monitor General Emergency Main Stack RAGEMS 8.69 E+01 uCi/cc HRM OR 3.53 E-08 amps HRM Turbine Building RAGEMS 7.17 E-01 uCi/cc HRM HRM = High Range Monitor 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 mRem while the 5000 mRem, thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE... .. ~ ~ ~ ~~~~~~~~~~~~~~..

..... .. ..':.. .., ... .". ... ..". ** * "- ...-"" Basis Refe rence(s)..:

1. NEI 99-01 Rev 6, AG1 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek 5. EP-AA-1 10-200, Dose Assessment

6. EP-AA-1 10-201, On Shift Dose Assessment xxxMonth2015 OCGS 3-34 EP-AA-1010 Rev XX r% afgar (rngak 1.dandarnfin Afnfinn Annoy I:=Ygxlfn IN1 A'.,a+car t~rn~I f~n~r~tinn

~t~atir~n Anncv FvMnn hi. irI~~r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RS1 lnitiating Condition:.........

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

... ......................

1,2,3,4, D Em.ergepncy Action Level (EAL);: ..... ..... ..Notes: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time.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. Readings on ANY Table R1 Effluent Monitor > Table RI value for> 15 minutes.OR 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER: a. > 100 mRem TEDE OR b. > 500 mRem CDE Thyroid OR 3. Field survey results at or beyond the site boundary indicate EITHER: a. Gamma (closed window) dose rates >100 mR/hr are expected to continue for > 60 minutes.OR b. Analyses of field survey samples indicate > 500 mRemCDE Thyroid for 60 minutes of inhalation.

xxxMonth2015 OCGS 3-35 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster..C..ek GeeaiaSainAnxEeo ula Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RSl (cont)Emerg ency Action:Lebe (EAL) (cant): Table RI Effluent Monitor Thresholds Effluent Monitor Site Area Emergency Main Stack RAGEMS 8.69 E+00 uCi/cc HRM OR 3.53 E-09 amps HRM Turbine Building RAGEMS 3.65 E+05 cpm LRM HRM = High Range Monitor LRM = Low Range Monitor 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 mRem while the 500 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Escalation of the emergency classification level would be via IC RG1.Basis. Reference(s):

1. NEI 99-01 Rev 6, AS1 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek 5. EP-AA-1 10-200, Dose Assessment

6. EP-AA-1 10-201, On Shift Dose Assessment xxxMonth2015 OCGS 3-36 EP-AA-1010 Rev XX I=xnlnn Nuclear r) cfar rtraak r-anarnting Statinn Annex Exelon Nuclear X Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA1 Initiating Condition:..

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mRem TEDE or 50 mRem thyroid CDE.'O~perating-Mode Applicability:

... ... .... ..... ..* ... ..... .... .............. .L ... ... ...... ... ........ .... .. '.. .: ....... .. ..1,2,3,4, D Emergency Action Level (EAL): ..Notes: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time." 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. 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 xxxMonth2015 OCGS 3-37 EP-AA-1 010 Rev XX C) cfar rratak rZanarntin Atnfinn Annoy Pviylnn Kiiiil-,nr flut~ (~~~ (0n~~tnn~t~in An~vFvIn MIr:32 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA1 (cont)Emergency Action Level (EAL). (.ont): 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 RI Effluent Monitor Thresholds Effluent Monitor Alert Main Stack RAGEMS 8.69 E-01 uCi/cc HRM OR 3.53 E-10 amps HRM Turbine Building RAGEMS 3.65 E+04 cpm LRM HRM = High Range Monitor LRM = Low Range Monitor 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 mRem while the 50 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.The radwaste liquid discharge system is currently closed off with a plant modification installed blank flange. To perform a discharge would require a plant modification to remove the flange. Since the liquid radwaste system is not operable, no EAL threshold has been developed for this release point.Escalation of the emergency classification level would be via IC RS1.xxxMonth2015 OCGS 3-38 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA1 (cont)Basis Reference(s):

1. NEI 99-01 Rev 6, AA1 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek.5. ABN-27, Inadvertent Overboard Radioactive Release or Cross Contamination

6. EP-EAL-0617, Oyster Creek Criteria for Choosing Liquid Effluent EAL Threshold Values xxxMonth2015 OCGS 3-39 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster Creek Generatina Station Annex Exelon Nuclear Tn lg f'rr-1 1-7V OCGS EAL Technical Basis w * ... .. lRECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RUI In11itiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.Operating Mode Applicability:-

... ... .. .. ..... ..-.. .. .... .. ............ .. ... ..... ...... ...,... .: ....." .' .*1,2,3,4, D Em~ergency Action Level, (EAL): Notes: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded." If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the applicable time." 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. Readings on ANY Table R1 Effluent Monitor > Table R1 value for > 60 minutes: Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Event Main Stack RAGEMS 4.07 E+03 cps LRM Turbine Building RAGEMS 4.16 E+02 cpm LRM LRM = Low Range Monitor OR 2. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.xxxMonth2015 OCGS 3-40 EP-AA-1 010 Rev XX FypInn NHr.lenr flvQtar C~reek (r0narztinn Atntinn Annex FvieInn Nuclea~r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RUl (cont)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.

Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.The radwaste liquid discharge system is currently closed off with a plant modification installed blank flange. To perform a discharge would require a plant modification to remove the flange. Since the liquid radwaste system is not operable, no EAL threshold has been developed for this release point.EAL #1 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous effluent pathways.EAL #2 Basis 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 RAI.xxxMonth2015 OCGS3-41 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RU1 (cont)Basis Reference(s):...

1. NEI 99-01 Rev 6, AU 2. EP-EAL-0610, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Oyster Creek Generating Station 3. BNE Correspondence dated February 1, 2007 4. CY-OC-1 70-301, Offsite Dose Calculation Manual for Oyster Creek 5. ABN-27, Inadvertent Overboard Radioactive Release or Cross Contamination xxxMonth2015 OCGS3-42 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RG2 Initiating Condition:

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

... ... .........1,2,3,4, D Emergency Action Level (EAL)::' ..... ....1. .... ... ... ... .. .. .. ...... ........ ... ... .... ...... .... .. .....Note: The Emergency Director should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for 60 minutes or longer.basis:, This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

ýBasis Reference(s)::

1. NEI 99-01 Rev 6, AG2 xxxMonth2015 OCGS3-43 EP-AA-1010 Rev XX F~xcdnn NH.ieInn~r tl Qfar rtrpak ranarging Station nnex Exelon Nuclear.L Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RS2 I)nitiating Condition:

Spent fuel pool level at (site-specific Level 3 description).

OpratingMode Applicability; 1,2,3,4, D Emergency Action Level (EAL):-Lowering of spent fuel pool level to (site-specific Level 3 value).... s -.. ..... --_ _ .. ...... .... .: = =: .: : :. ....: .: .......... ..... .............. , L .= ... : z This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC RG1 or RG2.Basis Reference(s):

1. NEI 99-01 Rev 6, AS2 xxxMonth2015 OCGS3-44 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA2 Initiating Condition;,., Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability;,,.

.1,2,3,4, D Emrerency Action Level (EAL): 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 mRem/hr.OR 3. Lowering of spent fuel pool level to (site specific Level 2 value).Table R2 Refuel Floor ARM's* C-5, Crit Mon* C-10, North Wall 0 C-9, North Wall* B-9, Open Floor B a s i s :........

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

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

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

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

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.xxxMonth2015 OCGS3-45 EP-AA-1010 Rev XX Exelon Nuclear flvstor Creeck Generatinn Station Annex ExeIon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA2 (cont).Basis (cont): S ' '... ...... .. .. ............ .. ....... .... ...... ... .... ... ..... .... .... .. ..... ... ...,.. ..... ... ....This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.EAL #1 Basis This EAL escalates from RU2 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 a rise 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 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).

EAL #3 Basis: Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Escalation of the emergency would be based on either Recognition Category R or C ICs.xxxMonth2015 OCGS3-46 EP-AA-1010 Rev XX rl afckr r-rank (Zimmmr2fin -Qfnfitn Annav IPvalnn Ikh Aanlr A~w4~r (~r~aIe t~nar~tinn

~t~tu~n Annav Fv~Inn M.Ir~I~2r

0 t 5 ES SeE *SeE ...... SI t&5E 5*

rs. *

Efl * ***Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA2 (cont)Basis Reference(s):

.1. NEI 99-01 Rev 6, AA2 2. RAP G-7-a, SKM SRG TNK LVL LO-LO 3. RAP-10F-I-m, Crit Mon C5 Hi 4. RAP-10F-3-m, North Wall C9 Hi Vent Trip 5. RAP-10F-2-m, North Wall C10 Hi 6. RAP-1i F-4-m, North Wall B9 Hi Vent Trip xxxMonth2015 OCGS 3-47 EP-AA-1010 Rev XX n fdr f"rAmMe elmnnrnfin Qfmfirn AnnAm-z I=,-,I~rhnn Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RU2 Initiating Condition:

--UNPLANNED loss of water level above irradiated fuel.P ... .. ~ ~ .. ... .... .. ..... .........

.... * ,, , :. ,, ..... ....... .. ...........

.... .... ..., Operating M ode Applicability:

1,2,3,4, D Emergency Action Level (EA L):.'1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

Refueling Cavity water level < 583 inches (GEMAC Wide Range, floodup calibration).

OR* Indication or report of a drop in water level in the REFUELING PATHWAY.AND b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R2.Table R2 Refuel Floor ARM's* C-5, Crit Mon* C-10, North Wall* C-9, North Wall" B-9, Open Floor xxxMonth2015 OCGS 3-48 EP-AA-1010 Rev XX l:xeInn n Qtar (rrook ronarntinn Rtntinn Annoy ExeInn Nui~clear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RU2 (cont)B a i : ....... ... :........

..... ..... .. ... ... .. ........ .. .... ..... .. .... ...I ..........

... ...Basis: UNPLANNED:

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

The cause of the parameter change or event may be known or unknown.REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.This IC addresses a loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level 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 a rise 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 rise 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 RA2.T .. .... .......... -_.. ..._., ..... ....._ I.- I .., .. .... ...: ....-.1. ..I... ..... -, ,, -I. ..... ...... .. .... ...* ... .. .......:. -.. *7... :...Basis Reference(s):.

1. NEI 99-01 Rev 6, AU2 2. RP-AA-203 Exposure Control and Authorization

3. RAP-G-7-a, SKM SRG TNK LVL LO-LO 4. 205.94.0 RPV Floodup Using Core Spray 5. 205.95.0 Reactor Flood-up / Drain-down

6. FSAR Figure 7.6-3 xxxMonth2015 OCGS3-49 EP-AA-1010 Rev XX Ovater Creek Generatina Station Annex Exelon Nuclear Oyster Creek Generatina Station Annex Exelon Nuclear i~hl C)MRVV VI-OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA3 Initiating Condition:

Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.p ra in .M o e ~ p ica iiiy ! .............

.. ............

..... ... ... .: .... .. ...... "" .........

Operating .Mode Applicability:, .. ..........

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

.. ... ......Note: If the equipment in the 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. Dose rate> 15 mR/hr in ANY of the following Table R3 areas: Table R3 Areas Requiring Continuous Occupancy* Main Control Room* Central Alarm Station -(by survey)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 51' elevation Modes 3 and 4 xxxMonth2015 OCGS 3-50 EP-AA-1010 Rev XX t-1 afar rragnk (~anarntinn Afnfinn AnnayF~n irIzr t Er .. Pyininn N"e-lonr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 (cont)Basils 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.

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.

Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR, therefore Table R4 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in service. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding beyond that required by procedures, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).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 rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.xxxMonth2015 OCGS 3-51 EP-AA-1010 Rev XX IExnlnn Nu.lrn~zr f) Qtor rrook Ganarnting Station nnex Exellon Nuclear.L Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENT RA3 (cont).B.as cis (cant): ... .........-..

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 R, C or F ICs.Basis Reference(s)::,-:

/-. .1. NEI 99-01 Rev 6, AA3 2. ABN-29, Plant Fires 3. EMG-3200.1 1, Secondary Containment Control Safe Shutdown Area xxxMonth2015 OCGS 3-52 EP-AA-1 010 Rev XX I=xelon Nuclear (hster C~reek Genncratinn Station Annex Exelon Nuclea~r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU3....... ... " .. "'. ". ..... ""i. .. ... ;.... ...... .. R" .U~ .3.. ...Initiating Condition:, Reactor coolant activity greater than Technical Specification allowable limits.i e. r a n ...... A p.i a b li.- .... ... ................

i ....... ... .. .... .... ....... ... ..... ..... .. ... ... ................

..... ...Oerating Mode Applicability:

1,2 Emergency ActionLevel (IAL): 1. Offgas system radiation monitor Hi-Hi alarm.OR 2. Specific coolant activity > 4.0 uCllgm Dose equivalent 1-131.Basis: ......... .This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.

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

'radiation monitor readings that provide indication of a degradation of fuel clad integrity.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.Basis ference .). ..... .1. NEI 99-01 Rev 6, SU3 2. Technical Specifications 3.6.A 3. ABN-26, High Main Steam Line or Off Gas Activity 4. RAP1OF-1-c, Offgas HI-HI xxxMonth2015 OCGS 3-53 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initating Cond~ition:

-Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier..Operating M. oe Applicability

.. .....:. ..1,2 EEmergency 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.Basisi Referenc(:es):

.1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-54 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 iln i ia ti g : C n d i o n :i-i ... .. ... .iii : ... : ll .........

.- ....... ..- ... ....... iu -i'....I.nli.tia-ti ng.cC~on~ditiio~n.:

Loss or Potential Loss of ANY two barriers.Operating Mode Applicability:

1,2 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-2 xxxMonth2015 OCGS 3-55 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS 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: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.B s Refe..en es). .. .. .. Table 9-F-2 1. NEI 99-0 1 Rev 6, Table 9-F-2 xxxIVlonth2015 OCGS 3-56 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FCl Inti ating. Condition:

.. .: ., .......... .., , .., ... .... .: .. RCS Activity Operating M.ode Applicability:

1,2 Fission ProductBarrier (FPB) Threshold:

LOSS Coolant activity > 300 uCi/gm Dose Equivalent 1-131.Bas is: This threshold indicates that RCS radioactivity concentration is greater than 300 jCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete.

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

There is no Potential Loss threshold associated with RCS Activity.Basis, Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-57 EP-AA-1 010 Rev XX t'% afmr Prgagale flnngarnfin Qf!afun Anngiv Nlliilior (~ r~~L (~aui~r~4inri

~4~*i,%n A nn~v ~v~Ir~n Mi ir~I~i~r*O t~. * ~ fl

S Sil * ..i 55 t* *

F1* * * *'..fl .S~ U *

I * .*Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:

RPV Water Level O.pe.atirng.Mod.-

App'licabiiy 1,2 F4ission Product Ba'rrier.(FPB)

Thres hold: LOSS 1. Plant conditions indicate primary containment flooding is required.POTENTIAL LOSS 2. RPV water level cannot be restored and maintained

> 0 inches TAF.OR 3. RPV water level cannot be determined.

Loss Threshold

1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.This is identified in the BWROG EPGs/SAGs when the phrase, "Primary Containment Flooding Is Required," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "Primary containment flooding required," also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.

Potential Loss Threshold

2 and #3 Basis This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.The RPV water level threshold is the same as RCS Barrier RC2 Loss threshold.

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

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

EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required xxxlVlonth2015 OCGS 3-58 EP-AA-1010 Rev XX I=xelon Nuclear~vhcmar Cr.ek Gnertinn Station Annex Ex,.on Nucle.ar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)Basis (cont):.?.-

...1.1 .- .1 ; 4.. -. ----...... .. ....... .. .. ..* ... ... .emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.

For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

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

Basis Reference(s):., 1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.01A, RPV Control -No ATWS 3. EMG-3200.01 B, RPV Control -With ATWS 4. EMG-3200.08A, RPV Flooding -No ATWS 5. EMG-3200.08B, RPV Flooding -With ATWS 6. EMG-3200.02, Primary Containment Control xxxMonth2015 OCGS 3-59 EP-AA-1010 Rev XX Nuclear Ovstcer Creek Geinersatinn S~tation Annex ExeIon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condilti~onR:.ý, .........Primary Containment Radiation Operating Mode Applicability:

1,2 Fission Product Th.reshold:

LOSS Containment Hi Range Radiation Monitoring System (CHRRMS) reading > 530 R/hr.B a s is : .'..'. ........ .....The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 ýiCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for RCS Barrier RC5 Loss Threshold 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.

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

,,Basis R eference(s'):

.-...........

........ ....-----1. NEI 99-01 Rev 6, Table 9-F-2 2. Core Damage Assessment Methodology xxxMonth2015 OCGS 3-60 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7-initiating Condition:

Emergency Director Judgment.OperatingaodeAppcabili.............

1,2 Fission Product Barrier (FPB) Threshold:

LOSS 1. Any condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.POTENTIAL LOSS 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.Basisi:'Loss Threshold

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

2 Basis 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

.1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-61 EP-AA-1 010 Rev XX Fxolnn Nuclear Ov--far Creek (G0neratinn Sta~tion Annexv Exelon NucleIAar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:

RPV Water Level Ope ratin-g'Mode-Aplcbiiy" 1,2 Fission Product Threshold:

LOSS 1. RPV water level cannot be restored and maintained

> 0 inches TAF.OR 2. RPV water level cannot be determined.

B asis:......

... ...This water level corresponds to the Top of Active Fuel (TAF) and is used in the EOPs to indicate challenge to core cooling.The RPV water level threshold is the same as Fuel Clad Barrier FC2 Potential Loss threshold.

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

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.xxxMonth2015 OCGS 3-62 EP-AA-1010 Rev XX l::xeInn Ovsnter Creek Generatinn Station Annexv Fvnlnn Nuclerkr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)B asis (cont) .............

...... ......In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.

For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

There is no RCS Potential Loss threshold associated with RPV water level.!Basis Refere"nce(s):

1. NEI 99-01 Rev 6, Table 9-F-2 2. 2000-GLN-3200.01, Plant Specific Technical Guideline 3. 2000-BAS-3200.02, EOP Users Guide xxxMonth2015 OCGS 3-63 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 initiating Condition:.......

Primary Containment Pressure Operating, ModeApplicability:

..... ... '.. .. ..1,2* ; .. .... ; .. .... .........o ... .... ... .. .... .. ..... ........ .... ................ ... ......Fission Product Barrier (FPB).Threshold:

LOSS 1. Drywell pressure > 3.0 psig.AND 2. Drywell pressure rise is due to RCS leakage Basis: ........The > 3.0 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure.

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

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

A stuck-open Electromatic Relief Valve (EMRV) or EMRV leakage is not considered either identified or unidentified leakage by Technical Specification and, therefore, is not applicable to this EAL.There is no Potential Loss threshold associated with Primary Containment Pressure.Basis Reference(s):

.......1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.01A, RPV Control -No ATWS 3. EMG-3200.02, Primary Containment Control 4. 2000-BAS-3200.02, EOP User's Guide xxxMonth2015 OCGS 3-64 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating

.ond ti ... ....o...... ... ......RCS Leak Rate Operating Mode Applicability:

1,2 Fission Product Barrier (FPB)JThreshold:

'LOSS 1. UNISOLABLE Main Steam Line (MSL), Isolation Condenser, Feedwater, or RWCU line break.OR 2. Emergency RPV Depressurization is required.POTENTIAL LOSS 3. UNISOLABLE primary system leakage that results in EITHER of the following:

a. Secondary Containment area temperature

> EMG-3200.11 Max Normal (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11 Max Normal (Table 12) operating level.Basis: UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or, locally.Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Classification of a system break over system leakage is based on information available to the Control Room from the event. Indications that should be considered are:* Reports describing magnitude of steam or water release.* Use of system high flow alarms / indications, if available," Significant changes in makeup requirements," Abnormal reactor water level changes in response to the event.The use of the above indications provides the Control Room the bases to determine that the ongoing event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.xxxMonth2015 OCGS 3-65 EP-AA-1010 Rev XX t'l afar ('rook rZanarnfinn Afnfinn nnay A~,~tr (~o~Ie(~n~~tinn~tzainn nnExFvlon Nucle~~ar Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont),Basis (c n):.........

K Loss Threshold

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

If it is determined that the ruptured line cannot be promptly isolated, the RCS barrier Loss threshold is met.Loss Threshold

2 Basis Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open Electromatic relief valves (EMRVs) and keep them open. Even though the RCS is being vented into the torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.Potential Loss Threshold

3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as Trunnion room, Isolation Condenser, RWCU etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification.

A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials.

Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold#1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.xxxMonth2015 OCGS 3-66 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)Basis Reference(s):.

1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.1 1, Secondary Containment Control 3. 2000-GLN-3200.01, Plant Specific Technical Guideline xxxMonth2015 OCGS 3-67 EP-AA-1010 Rev XX Oyster Creek Generatinq Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:........

' -Primary Containment Radiation Operating Mode Applicability:

1,2 Fission Prod uctBarrie!r(FPB)

Threshold:

LOSS Containment Hi Range Radiation Monitoring System (CHRRMS) reading > 10OR/hr.B..asis:.The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier FC5 Loss Threshold since it indicates a loss of the RCS Barrier only.There is no Reactor Coolant System Potential Loss threshold associated with Primary Containment Radiation.

Bpasi-kRefeec~)

-..-.. ... ...... ........ .... ... ... ..-..... ' ... ......... .. ...'= ' ....i il : .......1. NEI 99-01 Rev 6, Table 9-F-2 2. EP-EAL-061 1, Criteria for Choosing Containment Radiation Monitor Reading Indicative of Loss of RCS Barrier xxxMonth2015 OCGS 3-68 EP-AA-1010 Rev XX Exelon Nuclear Ovster C~reeck Genceratinn Station Annex ExeIon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:

.. .... ... ......... .. ....i ..: ...... *. ....." ....° .. :: .i ..... ....Emergency Director Judgment.Operating Mode Appicdab-ility:

..... ... ......1,2 F m s~~~o n p r~~~~d uc.t....r e (F B. .h r s h1. .... ..........

. .... i. :. _.-.. ...... _.:..._ :.'Fission Pro-duct Ba rr~iei (FýP.B)_Threshold:.

LOSS 1. Any condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier.POTENTIAL LOSS 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier..... " .:" .".'"" "". :" " .'" ." .' T .."" .. " * = " =' .."..".. .. * ...B a s is :: ...... .... ..... .. .... ....... .... ..... .... ...Loss Threshold

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

2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

B as~~~~~~~~~~~~~s~.. .ee e c ~ ) *::... ... ., .... .. ....... .............bas is Re9fe-r R eence s)1. NEI 99-0 1 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-69 EP-AA-1010 Rev XX I=xAInn Nuclear (Ivyar C~reek Generatinn S~tation Annex Exelnn Nucler~I Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating'C'o-n"d'itiobn:'-7

i n it ia i ng....t io n .. ....... -...... -.... .......:: ' .: ' .... .. ..... .....i , , : .... == = : = ,== ...RPV Water Level 0Operating Mode Applicability:

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

POTENTIAL LOSS Plant conditions indicate primary containment flooding is required.BaSis':'.

..The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV water level. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible.

BWR EOPs/SAMGs specify the conditions that require primary containment flooding.

When primary containment flooding is required, the EOPs are exited and SAMGs are entered. Entry into SAMGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 2. EMG-3200.01 B, RPV Control -With ATWS 3. EMG-3200.08A, RPV Flooding -No ATWS 4. EMG-3200.08B, RPV Flooding -With ATWS 5. EMG-3200.02, Primary Containment Control 6. EMG-3200.01A, RPV Control -No ATWS xxxIVIonth2015 OCGS 3-70 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Inritiating Condition:

Primary Containment Conditions O perating.

M ode Applicability:,.-

.. .... ....- ...... ... ... ........1,2 Fission Product Barrier (FPB)!.: hreshhold:

LOSS 1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.OR 2. Drywell pressure response not consistent with LOCA conditions.

POTENTIAL LOSS 3. Drywell pressure > 44 psig and rising.OR 4. a. Drywell or Torus Hydrogen concentration

> 6%.AND b. Drywell or Torus Oxygen concentration

> 5%.OR 5. Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded......... ....,.': ... .. " ... ..........

..... ......... ' .... ..... " .." " " -: ....... =... .:B asis:......=

.... * * .UNPLANNED:

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

The cause of the parameter change or event may be known or unknown.Loss Threshold

1 and #2 Basis Rapid UNPLANNED loss of Drywell pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of Drywell integrity.

Drywell pressure should rise as a result of mass and energy release into the Drywell from a LOCA. Thus, Drywell pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned.

The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

A pressure suppression bypass path would not be an indication of a containment breach.xxxlVlonth2015 OCGS 3-71 EP-AA-1 010 Rev XX Ovstar Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)Basis (cont): .. ............

Potential Loss Threshold

3 Basis The threshold pressure is the Drywell internal design pressure.

Structural acceptance testing demonstrates the capability of the Drywell to resist pressures greater than the internal design pressure.

A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.Potential Loss Threshold

4 Basis If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.Potential Loss Threshold

5 Basis The HCTL is a function of RPV pressure, torus temperature and torus water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

Basis Reference(s):

.... .....1. NEI 99-01 Rev 6, Table 9-F-2 2. FSAR Update 6.2.1.1.3 3. Technical Specifications

5.2 Basis

4. EMG-3200.02 Primary Containment Control xxxMonth2015 OCGS 3-72 EP-AA-1 010 Rev XX FxeInn Ovster C~reeik Gaeneratinn Sta~tion Anncx Fxelnn Nuclea~zr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 nitiating Condition:.

Primary Containment Radiation Operating Mode Applicability:

..1,2 Fission Product Barrieri FPBiTh'reshold:

'POTENTIAL LOSS Containment Hi Range Radiation Monitoring System (CHRRMS) reading > 1210 R/hr.Basis: There is no Loss threshold associated with Primary Containment Radiation.

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, 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-2 2. Core Damage Assessment Methodology (CDAM)xxxMonth2015 OCGS 3-73 EP-AA-1 010 Rev XX Ovater Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:

........ .: ..... ..Primary Containment Isolation Failure:Operating Mode Applicability:_

..........

...........

..1,2 Fission Product Barrier (FPB).ThrFeslhold:..

LOSS 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.OR 2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to accident conditions.

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

a. Secondary Containment area temperature

> EMG-3200.11 Max Safe (Table 11) operating level.OR b. Secondary Containment area radiation level > EMG-3200.11 Max Safe (Table 12) operating level.Basis: UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss Threshold

1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment.

Examples include unisolable Main Steamline, Isolation Condenser line breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.xxxMonth2015 OCGS 3-74 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)Basis (cont): 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.Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components.

Minor releases may also occur if a primary containment isolation valve(s)fails to close but the primary containment atmosphere escapes to an enclosed system.These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.Loss Threshold

2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded.

Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.

Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the Drywell high pressure scram setpoint) does not meet the threshold condition.

Loss Threshold

3 Basis The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.

EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharging inio the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials.

Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.xxxMonth2015 OCGS 3-75 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)Bsis ,(cont) ..In combination with RCS Barrier RC4 Potential Loss Threshold

3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.'Basis Reference(s):, ........ .... ..1. NEI 99-01 Rev 6, Table 9-F-2 2. 2000-GLN-3200.01, Plant Specific Technical Guideline 3. EMG-3200.02, Primary Containment Control 4. Support Procedures

-32, -34, -41, -44 5. 2000-GLN-3200.03, OCNGS Plant Specific Technical Guidelines for Severe Accident Guidelines

6. EMG-3200.1 1, Secondary Containment Control xxxMonth2015 OCGS 3-76 EP-AA-1010 Rev XX rl a* a ('an (Lo ani t*

t*

r-t. S IPvalrn IkJl"Mnr A~,c4~r (~ri~Ie (~nar2tinn

~t2tinn Ann~v Fv~Inn Muir~I~2r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:

Emergency Director Judgment.Ope.rating Mode Applicablity

...............

1,2ýFission Pro"duct Barrier (FPB) Threshid:'

LOSS 1. Any condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier.POTENTIAL LOSS 2. Any condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier..... s. .. .. .. .... ........ ............ .... ........ ....." ......... .." ............... ... ... ...... ............ ... ...........* .... ...... .. -...... .... .... ....... ..... ........ .:;.. .. ................ ...........

.. ... ... "; -- ---- --- -- -Loss Threshold

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

2 Basis This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

.... .... ....... .. ....... C-..- --. ..-.-.. ...;t ....!.., ........ .........: ! ." .: : 1Basis Reference(s):

...1. NEI 99-01 Rev 6, Table 9-F-2 xxxMonth2015 OCGS 3-77 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1.Initiating Condition:-, Prolonged loss of all Off-site and all On-Site AC power to emergency busses.operating Mode Applicability:

1,2 Emergency Action leVel (EAi): .. .Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1 Loss of ALL offsite AC power to 4160V Buses 1C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses IC and 1D.AND 3. EITHER of the following:

a. Restoration of at least one 4160V Bus (IC or 1 D) in < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is not likely.OR b. RPV water level cannot be restored and maintained

> -20 inches TAF.:Z , ;i 17 i i i I ...... .... ..... ........ ....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 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 any fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that. power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and xxxMonth2015 OCGS 3-78 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)Bas'is (cont): event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation.

Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.Basis .Reference(s):..

1. NEI 99-01 Rev 6, SGI 2. UFSAR Section 8.2, Offsite Power System 3. ABN-37, Station Blackout 4. ABN-60, Grid Emergency 5. Regulatory Guide 1.155, Station Blackout 6. TDR-1 099, "Station Blackout Evaluation Report" 7. 2000-BAS-3200.02, EOP User's Guide 8. 2000-GLN-3200.01, Plant Specific Technical Guideline 9. OCNGS Drawing BR 3000 10. ABN-36, Loss of Off-Site Power xxxMonth2015 OCGS 3-79 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 Initiating Condition:, Loss of all offsite and all onsite AC power to emergency busses for 15 minutes or longer..... .... ....... ... ..... ...... .. .Operkting Mode Applicability:

1,2;Em~ergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of ALL offsite AC Power to 4160V Buses 1 C and 1 D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D.AND 3. Failure to restore power to at least one 4160V Bus (1 C or 1 D) in < 15 minutes from the time of loss of both offsite and onsite AC power.B a s is :.*....*-...:*.*...*.

....SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via ICs RG1, FG1, MG1, or MG2.xxxMonth2015 OCGS 3-80 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)Basis Reference(s).:

_: 1. NEI 99-01 Rev 6, SS1 2. UFSAR Section 8.2, Offsite Power System 3. OCNGS Drawing BR 3000 4. ABN-36, Loss of Off-Site Power 5. ABN-37, Station Blackout 6. ABN-60, Grid Emergency xxxlVlonth2015 OCGS 3-81 EP-AA-1010 Rev XX fl a+dar frnnk 11nnnrnfin -q+2+ien Anneky Iwailnn N ra i nh~r Z~4~ V2ra,(n~~in~2ti~

nvF~n arI Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.,pebrating Mode. Aplicability 1,2 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. AC power capability to 4160V Buses 1C and 1 D reduced to only one of the following power sources for > 15 minutes.* Startup Transformer SA* Startup Transformer SB* EDG-1 Emergency Diesel Generator* EDG-2 Emergency Diesel Generator AND 2. ANY additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

This IC provides an escalation path from IC MUI.An "AC power source" is a source recognized in ABNs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

xxxMonth2015 OCGS 3-82 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)Basis (cont): " 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." 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 MS1.Basis Referenes-

1. NEI 99-01 Rev 6, SA1 2. UFSAR Section 8.2, Offsite Power System 2. OCNGS Drawing BR 3000 3. ABN-36, Loss of Off-Site Power 4. ABN-37, Station Blackout 5. ABN-60, Grid Emergency xxxlVlonth2015 OCGS 3-83 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU1 Initiating Condition:... Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability:

1,2 Emergency Action Level (EAL): Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Loss of ALL offsite AC power capability to 4160V Buses 1C and 1 D for >15 minutes.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.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 MAI.Basis Reference(s):.

... ...... A .... .1. NEI 99-01 Rev 6, SU1 2. UFSAR Section 8.2, Offsite Power System 3. OCNGS Drawing BR 3000 4. ABN-36, Loss of Off-Site Power 5. ABN-60, Grid Emergency xxxIVIonth2015 OCGS 3-84 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 initiating oCndt ion: ."." Loss of all AC and Vital DC power sources for 15 minutes or longer.O p e ~r tin g ~ d e =: p ~p i ;b il ty i-~i. ..: ..........

: :;........... .......".,:. ..... .... .........

.;.. .........

.... ......................

.Operating.

Mode, Applicability:-.

..... .....1,2 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of ALL offsite AC power to 4160V Buses1 C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D.AND 3. Voltage is < 115 VDC on 125 VDC battery busses B and C.AND 4. ALL AC and Vital DC power sources have been lost for > 15 minutes.B asiS ,: ..... .....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 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 all EAL conditions are met.xxxMonth2015 OCGS 3-85 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)Basis Reference(s).

1. NEI 99-01 Rev 6, SG8 2. UFSAR Section 8.3.2, DC Power Systems 3. UFSAR Section 8.2, Offsite Power System 4. OCNGS Drawing BR 3000 5. ABN-36, Loss of Off-Site Power 6. ABN-37, Station Blackout 7. ABN-60, Grid Emergency 8. ABN-54, Loss of DC Distribution Center B 9. ABN-55, Loss of DC Distribution Center C xxxlVlonth2015 OCGS 3-86 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2........ ......... .. ...: , ...... .... ....... .... .. ..,Initiating Conditon: Loss of all vital DC power for 15 minutes or longer.Operating Mode Applicability:

1,2 ,Emergency Action'Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Voltage is < 115 VDC on 125 VDC battery busses B and C for> 15 minutes.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 RG1, FG1 or MG2.Basis Reference(s):.

1. NEI 99-01 Rev 6, SS8 2. OCNGS Drawing BR 3000 3. ABN-54, Loss of DC Distribution Center B 4. ABN-55, Loss of DC Distribution Center C xxxMonth2015 OCGS 3-87 EP-AA-1 010 Rev XX Ovster Creek Generatinq Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 initiating Conditioni" Inability to shutdown the reactor causing a challenge to RPV Water Level or RCS heat removal.i~~~~~p e ~... ra in .......- ...p i a b ~ t .... .....................

..... ....... ... ..: ...... .. -........-*x , , -. .. ..... ..-,Operating Mode Applicability:,... .. ....... .........

.. .* , ,,,,

,* , ,, : :: ,;,,,'. ... .. ..... , , ,,. ,,,..,.,,...

.....-E.mergency iAction Level (EA: -1. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 2%.AND 3. EITHER of the following conditions exist:* RPV water level cannot be restored and maintained

> -20 inches TAF.OR* Heat Capacity Temperature Limit (EMG-3200.02 Fig. F) exceeded.:Basis:_This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection, are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.

This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria;EAL #3 is considered to be exceeded when, as specified in the site-specific EOPs, RPV water level cannot be restored and maintained above the specified level.Escalation of the emergency classification level would be via IC RG1 or FGI.xxxIVIonth2015 OCGS 3-88 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)Basis Reference(s;)-:

1. NEI 99-01 Rev 6, SS5 2. EMG-3200.01 B, RPV Control -with ATWS 3. EMG-3200.02, Primary Containment Control 4. 2000-BAS-3200.02, EOP User's Guide 5. 2000-GLN-3200.01, Plant Specific Technical Guideline 6. EMG-3200.01A, RPV Control -no ATWS xxxIVlonth2015 OCGS 3-89 EP-AA-1010 Rev XX t-1 a+dmr 1'radak I'Zonarnfin -Q+m+inn Anniav I::v,-Inn N, e-lior zu~a (n rtn : t2 t n n v v fh uaI2 L*

5 *5%* L**t **l***UU Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 Initiating Condition:.-

Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.Operating Mode Applica.bility:.........

A.l. ......I ,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. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 2%... :..... ... ... :... :. .. ... ....:. ..., ......... .; i. .. :. .: ...: ..,: ..: :. :.. ..: .. .,Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor consoles to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor consoles since this event entails a significant failure of the RPS.A manual action at the reactor consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor consoles (e.g., locally opening breakers).

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.xxxMonth2015 OCGS 3-90 EP-AA-1 010 Rev XX f) afar rrook lZanarnfin -Qfnfitn nngav I=valen KI"i-lonr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 (cont),Basis (cont): The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MS3. Depending upon plant responses and symptoms, escalation is also possible via IC FSI. Absent the plant conditions needed to meet either IC MS3 or FS1, an Alert declaration is appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Bsis -Reference(s):...........-

1. NEI 99-01 Rev 6, SA5 2. EMG-3200.01A, RPV Control -no ATWS 3. EMG-3200.01 B, RPV Control -with ATWS 4. 2000-BAS-3200.02, EOP User's Guide 5. 2000-GLN-3200.01, Plant Specific Technical Guideline xxxlVlonth2015 OCGS 3-91 EP-AA-1 010 Rev XX r) afgar ('rdank eZonckrnfin -qfnfitn nnow P::alnn Ni irlanr fliie4~r f~ra~Ie (~g~n~r2tinn

~*2+it'~fl Ann~v FvMnn kImir~I~2r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 Inititing Condition:

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

..........1 Emergenc-Y'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. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 2%.AND b. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.OR 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power> 2%.AND b. EITHER of the following:

1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.OR 2. Subsequent automatic scram / ARI is successful in shutting down the reactor.Basis:: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, xxxMonth2015 OCGS 3-92 EP-AA-1010 Rev XX rl a+ar (rimair (' Zanarnfin -q+2+ien Annav I:vnltn Khielanr z~aa Z2~nr4in~~in A nv~an irI~ 0.

5 ** 4fS*4 ~ *r.* .lf~ .***--*--Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)Basis.*(cnt):

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 take manual action at another location(s) on the reactor consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor consoles".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MA3. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC MA3 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied." If the signal generated as a result of plant work causes a plant transient that creates a real condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated." If the signal generated as a result of plant work does not cause a plant transient but should have generated an RPS scram signal and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

xxxMonth2015 OCGS 3-93 EP-AA-1010 Rev XX r) afmr (rngmle r-'Ammrn+in

-4Z+!m+i^n AnnAmv I~vg~lnn kllflarlr I'tor ks~~~ n 4trnAn~ v~n irIa Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)Basis Reference(s):..

1. NEI 99-01 Rev 6, SU5 2. EMG-3200.01A, RPV Control -no ATWS 3. EMG-3200.01 B, RPV Control -with ATWS 4. 2000-BAS-3200.02, EOP User's Guide 5. 2000-GLN-3200.01, Plant Specific Technical Guideline xxxIVIonth2015 OCGS 3-94 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis m I RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 Initiating Condition, UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mod.Appliabi!ty:

',. "...... ...... n gi .. .. ....L _ .. ...... .... .i:... ...1,2 Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. UNPLANNED event results in the inability to monitor ANY Table Ml parameter from within the Control Room for > 15 minutes.Table M1 Control Room Parameters 0 Reactor Power* RPV Water Level* RPV Pressure* Drywell Pressure* Torus Water Level* Torus Water Temperature AND 2. ANY Table M2 transient in progress.Table M2 Significant Transients

Turbine Trip* Reactor Scram* ECCS Actuation* Thermal power change > 25%* Thermal Power oscillations

> 10%xxxlVlonth2015 OCGS 3-95 EP-AA-1010 Rev XX f% efmr PrnnLe f2nnnrnfin Qf!%+itn nmav I:v,-lr,,n N, ,lanr I ~~ kita+4it nnvF lrnMua Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 (cont)'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 any 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-1022) to determine if an NRC event report is required.

The event would be reported if it significantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RS1.Basis Reference(s):

1. NEI 99-01 Rev 6, SA2 xxxMonth2015 OCGS 3-96 EP-AA-1 010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4:Initiating Condition:-- , ..UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability:

..........

.. .-. .......1,2 Emergency Action Level (EAL):: Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.UNPLANNED event results in the inability to monitor ANY Table M1 parameters from within the Control Room for > 15 minutes.Table M1 Control Room Parameters" Reactor Power* RPV Water Level* RPV Pressure* Drywell Pressure* Torus Water Level* Torus Water Temperature.........

~ ~ ~ .... ... ... ... ... ......... ... :. ....... ... : :,.. .. ...: ...... .. .. .. = ....UNPLANNED:

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

The cause of the parameter change or event may be known or unknown.SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.As used in this EAL, an "inability to monitor" means that values for any 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.xxxMonth2015 OCGS 3-97 EP-AA-1010 Rev XX t'l afdar PrAmnle 1--nnnrýfin Qf!2firn nndav A.,e4~r i'~ r~~L' (~n~r~4. nil ~4~+ir~n An n~v ~vaIr~n Mmuv~Ic~r

~L* -. ~ ~R *~I L** *t b~AL*,. fl~ ** *4 n... *~.A~D*L4*

Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4 (cont),Basis (cont);..:.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event would be reported if it significantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC MA4.B a is R fe e n ) i ! :i: :........

.. .. .. ..... ....... .: ..... .. .............

B1a-sis-Reference(s):

.s ...... ...... .........

... .......E. .... ..R. .....1. NEI 99-01 Rev 6, SU2 xxxMonth2015 OCGS 3-98 EP-AA-1 010 Rev XX Ovster Creak Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5.Initiating Condition::

Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.Operating Mode A pplicabillityf:--

1,2 Emergency Action Level (EAL): Note: If it is determined that the conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.1. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)" Internal or external flooding event" High winds or tornado strike" FIRE" EXPLOSION" Other events with similar hazard characteristics as determined by the Shift Manager AND 2. EITHER of the following:

a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.OR b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required 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 large 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 energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if the attributes of an explosion are present.xxxMonth2015 OCGS 3-99 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 (cont)Basis (cont):. ...... ....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.

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, 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 equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in 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 #2.b 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 operation or readily apparent through indications alone, 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 RS1.If the EAL conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.Ba~sis'Rfeenek)

1. NEI 99-01, Rev 6 SA9 2. ABN-38 xxxMonth2015 OCGS 3-100 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 iln ti ti g ~ n d ti n i -ii i I. ...... ........

.........

..i .... i ii il ..... M =Y=i.ii Initiating Condition:~

RCS leakage for 15 minutes or longer.Operating..

..de.Applica ility:. ..... .. .1,2 Emergency i:Action Level (EAL):.Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for> 15 minutes.OR 2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.OR 3. Leakage from the RCS to a location outside the Drywell >25 gpm for > 15 minutes.Basis: UNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

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 or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.

Lesser values typically require time-consuming xxxIVlonth2015 OCGS 3-101 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)Basi's (con): 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 any relief valve does not warrant an emergency classification.

A stuck-open Electromatic Relief Valve (EMRV) or EMRV leakage is not considered either identified or unidentified leakage by Technical Specification and, therefore, is not applicable to this EAL.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 ICs of Recognition Category R or F.Basis Reference(s):

1. NEI 99-01 Rev 6, SU4 2. Technical Specifications 3.3.D, Reactor Coolant System Leakage xxxMonth2015 OCGS 3-102 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Oyster..Creek....n.r...n.

Station. Ane..Ex..n.Nuc.ea Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7.Initiating Condition:_

Loss of all On-site or Off-site communications capabilities.

Operating Mode 1,2 Emergency Action Leve! (EAL): ..1. Loss of all Table M3 Onsite communications capability affecting the ability to perform routine operations.

OR 2. Loss of all Table M3 Offsite communication capability affecting the ability to perform offsite notifications.

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

Table M3 Communications Capability System Onsite Offsite NRC Plant Paging System X Station Radio X Conventional Telephone lines X X X Satellite Phone system X X NARS X HPN X X ENS X X;Basis: -'This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).xxxlVlonth2015 OCGS 3-103 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)Basis(cont)

...EAL #1 Basis Addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are listed in procedure EP-MA-1 14-100-F-03, State/Local Notification Form.EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU6 2. UFSAR Section 9.5.2, Communication Systems 3. EP-AA-1 24-1001, Facilities Inventories and Equipment Tests xxxlVlonth2015 OCGS 3-104 EP-AA-1 010 Rev XX Ovater Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1'Initiating Condition:

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

3,4, D Emergency Action Level.(EAL):.

Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. Loss of all offsite AC power to 4160V Buses 1C and 1D.AND 2. Failure of EDG-1 and EDG-2 Emergency Diesel Generators to supply power to 4160V Buses 1C and 1D..AND 3. Failure to restore power to at least one 4160V bus (1C or 1 D) in < 15 minutes from the time of loss of both offsite and onsite AC power.Ba.asis: ---. ..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.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level would be via IC CS6 or RS1.xxxMonth2015 OCGS 3-105 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA1 (cont)n ...... .... ....... ...... ..........

! 7 ... ...... ..... .....Basis ,Reference(sý):.

1. NEI 99-01 Rev 6, CA2 2. ABN-36, Loss of Off-Site Power 3. ABN-37, Station Blackout 4. ABN-60, Grid Emergency 5. OCNGS Drawing BR 3000 6. UFSAR Section 8.2, Offsite Power System xxxlVIonth2015 OCGS 3-106 EP-AA-1010 Rev XX C) Q+ar rrook Mnnarnfinn A+nfinn Annoy FvI:xlnn NI.ilI~sr Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU1 Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.O perating M ode Applicability

_:_. :: : .........

...... ..3,4, D Emergency Action Level (EAL): Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.1. AC power capability to 4160V Buses 1C and 1 D reduced to only one of the following power sources for > 15 minutes." Startup Transformer SA" Startup Transformer SB* EDG-1 Emergency Diesel Generator* EDG-2 Emergency Diesel Generator AND 2. ANY additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS...... ..............

.... ..... ... ... ...:... ...... .... ... ...'Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.xxxMonth2015 OCGS 3-107 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CUI (cont)initiating Condition:

.An "AC power source" is a source recognized in ABNs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below." A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator)." A loss of all 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." 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.The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA1.Basis- Reference(s):

B as s _R efe en...)..i

..... ..... ............

..._, L ,_1. NEI 99-01 Rev 6 CU2 2. ABN-36, Loss of Off-Site Power 3. ABN-37, Station Blackout 4. ABN-60, Grid Emergency 5. OCNGS Drawing BR 3000 6. UFSAR Section 8.2, Offsite Power System.xxxIVlonth2015 OCGS 3-108 EP-AA-1 010 Rev XX NHnlenr Ovster Creek Generatinn Station nnex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:

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

3,4 Ergency Action Level (EAL): ........Note: If it is determined that the conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.1. The occurrence of ANY of the following hazardous events: " Seismic event (earthquake)

Internal or external flooding event" High winds or tornado strike o FIRE" EXPLOSION" Other events with similar hazard characteristics as determined by the Shift Manager AND 2. EITHER of the following:

a. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.OR b. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required by Technical Specifications for the current operating mode.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 large 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 energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if the attributes of an explosion are present.xxxMonth2015 OCGS 3-109 EP-AA-1 010 Rev XX t'l afar rrank (' Zoniarnfin Afnfirn nnow I::valnn Ihlrlalnr Z',ca (:a~ ~n~~in~2tn nivF InN.ia0 Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA2 (cont)Basis(cont).:..

...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.

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, 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 equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in 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 #2.b 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 operation or readily apparent through indications alone, or 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 CS6 or RS1.If the EAL conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.B a is R ef er.e n c e (s): .... .... ... ..... ....... .I ... ...... .............-----

1. NEI 99-01 Rev 6, CA6 2. ABN-38, Station Seismic Event 3. FSAR Update Section 3. 7 (Seismic)4. FSAR Update Section 3.3.1 (High winds)5. ABN-31, High Winds 6. ABN-32, Abnormal Intake Level 7. ABN-29, Plant Fires xxxMonth2015 OCGS 3-110 EP-AA-1 010 Rev XX NHnlnnr Ovster C~reek (Generatinn Station Annex~ FvaeIn NuclerIzr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:!

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

3,4 Emergency Action Level (EAL):.Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.Voltage is < 115 VDC on required 125 VDC battery busses B and C for> 15 minutes.BSa sis: 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 operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.

For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Depending upon the event, escalation of the emergency classification level would be via IC CA6 or CA5, or an IC in Recognition Category R.Basis Refereince(s):

1. NEI 99-01 Rev 6, CU4 2. UFSAR Section 8.3.2, DC Power Systems 3. OCNGS Drawing BR 3000 4. ABN-54, DC Bus B and Panel/MCC Failures 5. ABN-55, DC Bus C and Panel/MCC Failures xxxMonth2015 OCGS 3-111 EP-AA-1010 Rev XX tl afar f'r~aak (Zanarnfinn qtntir AnnoyN Pyainn N"ellear A a Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU4 I1nitiating Condition:

Loss of all onsite or offsite communications capabilities.

3,4, D Emergency Action Level (EAL): 1. Loss of all Table C1 Onsite communications capability affecting the ability to perform routine operations.

OR 2. Loss of all Table C1 Offsite communication capability affecting the ability to perform offsite notifications.

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

Table Cl -Communications Capability System Onsite Offsite NRC Plant Paging System X Station Radio X Conventional Telephone lines X X X Satellite Phone System X X NARS X HPN X X ENS X X ,Basis:.......

This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).EAL #1 Basis Addresses a total loss of the communications methods used in support of routine plant operations.

xxxMonth2015 OCGS 3-112 EP-AA-1010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN REFUELING SYSTEM MALFUNCTIONS CU4 (cont)Basis (cont): EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are listed in procedure EP-MA-1 14-100-F-03, State/Local Notification Form.EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Basis Reference(s):

1. NEI 99-01 Rev 6, CU5 2. UFSAR Section 9.5.2, Communication Systems 3. EP-AA-1 24-1001, Facilities Inventories and Equipment Tests xxxIVIonth2015 OCGS 3-113 EP-AA-1010 Rev XX t'l czfnr ('rddile I'Znnjarnfin

_Q+nfinn nndav IPyah-in !M1, ,alr Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CA5 Initiating Cod'ition:-

Inability to maintain the plant in cold shutdown.O p e r a t m ~ ... -I .Ie........i.

... ... ... ..........

l. Il.. .. ..ll ll ll ll..l. l~l ll ~~ ll l~ ~~ l l. ... ..l5 ll.... ....bpe'raitin-g Model Applicability:

3,4 Emergency Action Level (EAL): .....Note: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.

1. UNPLANNED rise in RCS temperature

> 212*F for > Table C2 duration.Table C2 RCS Heat-up Duration Thresholds RCS Containment Closure Heat-up Status Status Duration Intact Not Applicable 60 minutes*Established 20 minutes*Not Intact Not Established 0 minutes* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, then EAL #1 is not applicable.

OR 2. UNPLANNED RPV pressure rise > 10 psig as a result of temperature rise.B a s is : ........ ...........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.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

xxxIVIonth2015 OCGS 3-114 EP-AA-1010 Rev XX r) atar rrack ('Zanarnfin Afnfirn nnay PyIn:Yln Kimlolr (httar (~raak t~c~ncr2tinn

~t2tinn Annav Fvalnn N.Ir4~2r Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)Basis(.(cont):

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise.The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature rise without a substantial degradation in plant safety.Finally, in the case where there is a rise in RCS temperature, the RCS is not intact, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).

This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.EAL #2 provides a pressure-based indication of RCS heat-up.Escalation of the emergency classification level would be via IC CS6 or RS1.Basis Refernce(s):..........

1. NEI 99-01 Rev 6, CA3 2. Technical Specifications 1.7 3. OU-AA-103, Shutdown Safety Management Program xxxMonth2015 OCGS 3-115 EP-AA-1010 Rev XX rl afar rragak rZanarnfin Afnfinn Annov A~,~ejr ~rc~i~e(~n~~tin L~t~tin An~ExFelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CU5=!n~~~~~~~~t~~~~at~~~~~ng~.. .o nd it o n....! .. ... .= ..: .. .... ... ....Initating Condition:-

UNPLANNED rise in RCS temperature

=Operating Mode Applicability:

3,4 Emer-gency Action Level (EAL): .Note: " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.

1. UNPLANNED rise in RCS temperature

> 212 0 F.OR 2. Loss of the following for >15 minutes." ALL RCS temperature indications AND* ALL RPV water level indications

.. .... ...... .... ... ...... ... .:... .... ...., * ..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.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA5.RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).xxxMonth2015 OCGS 3-116 EP-AA-1010 Rev XX Ovster Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 (cont)...s .....:.. ... ...... ...i ,i i~~ .:i i ... i , ...... .. ....: ... ...... ...... ...... ..i'Basis cont):-A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

EAL #1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.

During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.

A loss of forced decay heat removal at reduced inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown..EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA6 based on an inventory loss or IC exceeding plant configu ration-specific time criteria..B asis Reference(s):

.,.. ... .. .-... ...... .... .. ....1. NEI 99-01 Rev 6, CU3 2. Technical Specifications 1.7 3. 2000-GLN-3200.01, Plant Specific Technical Guideline 4. 205.94.0, RPV Floodup Using Core Spray 5. 205.95.0, Reactor Flood-up / Drain-down CA5 based on xxxlVlonth2015 OCGS 3-117 EP-AA-1 010 Rev XX Ovstar Creek Generatina Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 i~h~ti~ t n g

n d it o n i I : ::; ; : d ' : , ..... .. .- .................

..........initiating Condition:

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

0.perating Mo-de 'App icaibitity:-

O p ~~~.. in g .p .- l ....... .... .......... :...... ..... ... ..... ." ..... ..

.- ....... ...........

....... .... ...3,4 Emergency Action Level (EAL): -..._ .......Note:* The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.a. RPV water level < 0 inches TAF for > 30 minutes.AND b. Any Containment Challenge Indication (Table C4)OR 2. a. RPV water level cannot be monitored for > 30 minutes.AND b. Core uncovery is indicated by ANY of the following: " Table C3 indications of a sufficient magnitude to indicate core uncovery.OR* Refuel Floor Area Radiation Monitor C-10, North Wall, reading>3 R/hr.AND c. ANY Containment Challenge Indication (Table C4)Table C3 Indications of RCS Leakage* UNPLANNED floor or equipment sump level rise** UNPLANNED Torus level rise** UNPLANNED vessel make up rate rise** Observation of leakage or inventory loss*Rise in level is attributed to a loss of RPV inventory.

xxxMonth2015 OCGS 3-118 EP-AA-1010 Rev XX C) afar ('rook rZanarnfin Atnfinn nnayNHr.IAnr Au~hi (~r~~k ~~n~~tin ~t~inn nnExFvlIn Nui~clear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)Emergency Action Level (EAL) (cont): Table C4 Containment Challenge Indications

Primary Containment Hydrogen Concentration

> 6% and Oxygen Concentration

> 5%" UNPLANNED rise in containment pressure* CONTAINMENT CLOSURE not established*

Any Secondary Containment radiation monitor > EMG-3200.11 Maximum Safe (Table 12).* 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.B~asis: 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.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.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary for BWR) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.

This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA Protective Action Guidelines (PAG) exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.xxxMonth2015 OCGS 3-119 EP-AA-1 010 Rev XX Oyster Creek Generating Station Annex Exelon Nuclear Table OCGS 3-2 OCGS EAL Technical Basis RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)Basis: (cant):.UNPLANNED: