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| {{#Wiki_filter:ATTACHMENT 10 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR QUAD CITIES NUCLEAR POWER STATION EP-AA-1 006 Enclosures 6 Enclosure 10A - Revised EAL Comparison Matrix Document 0 Enclosure 10B - Revised EAL Red-Line Basis Documents 0 Enclosure 10C - Revised EAL Basis Documents | | {{#Wiki_filter:}} |
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| Justification 99-01 Proposed EAL I 99-01 Rev NEINEI Rev 66 Proposed EAL Justification Initiating Condition - GENERAL EMERGENCY AGI Initiating Condition:
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| RG1 H No Change Difference Deviation Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem Release of gaseous radioactivity resulting in offsite dose greater than 1,000 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 5,000 mrem thyroid CDE. mRem TEDE or 5,000 mRem thyroid CDE. classification.
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| Operating Mode Applicability: All Operating Mode Applicability:
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| : 2) Added the following to bullet #3 "Classification based on effluent monitor 1,2,3,4,5,D readings assumes that a release path to the environment is established." In order Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): 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 Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to
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| * The Emergency Director should declare the General Emergency promptly isolate the release path, then the effluent monitor reading is no longer valid for upon determining that the applicable time has been exceeded, or will likely be classification purposes." This allows for more timely classification since all the
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| * The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded.
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| that the applicable time has been exceeded, or will likely be exceeded. readily available on the 11x17 procedure matrix used by the SM.
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| * If an ongoing release is detected and the release start time is unknown,
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| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
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| assume that the release duration has exceeded 15 minutes.
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| I the effluent flow past an effluent monitor is known to have stopped due to if actions to isolate the release path, then the effluent monitor reading is no
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| * 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 longer valid for classification purposes.
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| known to have stopped due to actions to isolate the release path, then the
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| * The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.
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| used for emergency classification assessments until the results from a dose The pre-calculated effluent monitor values presented in EAL #1 should be assessment using actual meteorology are available.
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| used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
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| : 1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer: 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCilsec for > 15 minutes (as determined by Control Room Panels or PPDS (site specific monitor list and threshold values) - Total Noble Gas Release Rate).
| |
| OR
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| : 2. Dose assessment using actual meteorology indicates doses at or beyond the
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| : 2. Dose assessment actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose site boundary of EITHER:
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| receptor point) a. > 1000mRemTEDE OR
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| : 3. Field survey results indicate EITHER of the following at or beyond (site b. > 5000 mRem CDE Thyroid specific dose receptor point):
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| OR
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| * Closed window dose rates greater than 1000 mR/hr expected to 3. Field survey results at or beyond the site boundary indicate EITHER:
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| continue for 60 minutes or longer.
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| : a. Gamma (closed window) dose rates >1000 mR/hr are expected to
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| * Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.
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| 5000 mrem for one hour of inhalation.
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| OR
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| : b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.
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| Page 1 of 66
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| NEI 99-01 Rev 6 1 Proposed EAL Justification Initiating Condition - SITE AREA EMERGENCY AS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem Initiating Condition:
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| RS1 E No Change E Difference 1 Deviation Release of gaseous radioactivity resulting in offsite dose greater than 100 mRem 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 500 mrem thyroid CDE. TEDE or 500 mRem thyroid CDE. classification.
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| Operating Mode Applicability: All Operating Mode Applicability:
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| : 2) Added the following to bullet #3 " Classification based on effluent monitor 1,2,3,4,5,D readings assumes that a release path to the environment is established." In order Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): 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 Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to
| |
| * The Emergency Director should declare the Site Area Emergency promptly isolate the release path, then the effluent monitor reading is no longer valid for upon determining that the applicable time has been exceeded, or will likely be classification purposes." This allows for more timely classification since all the exceeded.
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| * The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor that the applicable time has been exceeded, or will likely be exceeded. readily available on the 1 x17 procedure matrix used by the SM.
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| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
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| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
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| * 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
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| * Classification based on effluent monitor readings assumes that a release path longer valid for classification purposes. 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
| |
| * The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.
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| used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. 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.
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| : 1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer:
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| : 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCl/sec for> 15 minutes (as determined by Control Room Panels or PPDS -
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| (site specific monitor list and threshold values) Total Noble Gas Release Rate).
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| OR
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| : 2. Dose assessment actual meteorology indicates doses greater than 1000 2. Dose assessment using actual meteorology indicates doses at or beyond the mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose site boundary of EITHER:
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| receptor point) a. >100 mRem TEDE OR
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| : 3. Field survey results indicate EITHER of the following at or beyond (site b. > 500 mRem CDE Thyroid specific dose receptor point):
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| OR
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| . Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer. 3. Field survey results at or beyond the site boundary indicate EITHER:
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| : a. Gamma (closed window) dose rates >100 mR/hr are expected to
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| . Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.
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| 500 mrem for one hour of inhalation.
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| OR
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| : b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.
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| Page 2 of 66
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| NEI 99-01 Rev 6 Proposed EAL Justification AA1 RAI Initiating Condition - ALERT Initiating Condition:
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| D No Change E Difference 1-1 Deviation Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 1) Listed site-specific monitors and Threshold values to ensure timely classification.
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| mrem TEDE or 50 mrem thyroid CDE. mrem TEDE or 50 mrem thyroid CDE.
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| Operating Mode Applicability: All Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1, 2, 3,4, 5, D to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): readings assumes that a release path to the environment is established, If the effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for
| |
| * The Emergency Director should declare the Alert promptly upon classification purposes." This allows for more timely classification since all the basis determining that the applicable time has been exceeded, or will likely be information pertaining to Note bullet 3 will be contained in the IC and therefor
| |
| * The Emergency Director should declare the event promptly upon determining readily available on the 1 lx1 7 procedure matrix used by the SM.
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| exceeded. that the applicable time has been exceeded, or will likely be exceeded.
| |
| * If an ongoing release is detected and the release start time is unknown,
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| * If an ongoing release is detected and the release start time is unknown, assume 3) A calculation was performed, in accordance with (lAW) guidance provided in NEI assume that the release duration has exceeded 15 minutes. that the release duration has exceeded 15 minutes. 99-01 revision 6 EAL AA1, to determine the effluent monitor response for a
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| * If the effluent flow past an effluent monitor is known to have stopped due radioactive liquid release via the normal site release pathway. The liquid release Classification based on effluent monitor readings assumes that a release path would contain activity equivalent to provide lmrem TEDE or 50mrem thyroid CDE to actions to isolate the release path, then the effluent monitor reading is to the environment is established. If the effluent flow past an effluent monitor is at the site boundary. The calculation determined the effluent monitor response no longer valid for classification purposes. known to have stopped due to actions to isolate the release path, then the would be >110% of the instrument maximum range and as such, lAW NEI 99-01
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| " The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes. Rev 6 guidance, was not included in this EAL.
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| used for emergency classification assessments until the results from a The pre-calculated effluent monitor values presented in EAL #1 should be used dose assessment using actual meteorology are available. for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
| |
| : 1. Reading on any of the following radiation monitors greater than the
| |
| : 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 reading shown for 15 minutes or longer: uCilsec for > 15 minutes (as determined by Control Room Panels or PPDS
| |
| - Total Noble Gas Release Rate).
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| OR (site-specific monitor list and threshold values)
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
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| : 2. Dose assessment actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose a. > 10 mRem receptor point) TEDE
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| : 3. Analysis of a liquid effluent sample indicates a concentration or release OR rate that would result in doses greater than 10 mrem TEDE or 50 mrem b. > 50 mRem CDE Thyroid thyroid CDE at or beyond (site-specific dose receptor point) for one hour of exposure. OR
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| : 3. Analysis of a liquid effluent sample indicates a concentration or release rate
| |
| : 4. Field survey results indicate EITHER of the following at or beyond (site that would result in doses greater than EITHER of the following at or beyond specific dose receptor point): the site boundary
| |
| * Closed window dose rates greater than 10 mR/hr expected to a. 10 mRem TEDE for 60 minutes of exposure continue for 60 minutes or longer.
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| OR
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| * Analysis of field survey samples indicate thyroid CDE greater than
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| : b. 50 mRem CDE Thyroid for 60 minutes of exposure 50 mrem for one hour of inhalation.
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| OR
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| : 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 k 60 minutes.
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| OR
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| : b. Analyses of field survey sam ples in dicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.
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| Page 3 of66
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| NEI 99-01 Rev 6 ] Proposed EAL Justification Initiating Condition - UNUSUAL EVENT AU1 Initiating Condition:
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| RUI D No Change E Difference 1 Deviation Release of gaseous or liquid radioactivity greater than 2 times the (site-specific Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 1) Listed site-specific monitors and Threshold values to ensure timely classification.
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| effluent release controlling document) limits for 60 minutes or longer 60 minutes or longer.
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| Operating Mode Applicability: All Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1, 2, 3,4,5, D to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3 Emergency Action Level (EAL): readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for
| |
| * The Emergency Director should declare the Unusual Event promptly upon
| |
| * The Emergency Director should declare the event promptly upon determining classification purposes." This allows for more timely classification since all the basis determining that 60 minutes has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded. information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded. readily available on the 1 x17 procedure matrix used by the SM.
| |
| * If an ongoing release is detected and the release start time is unknown, assume
| |
| * If an ongoing release is detected and the release start time is unknown, that the release duration has exceeded 60 minutes.
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| assume that the release duration has exceeded 60 minutes.
| |
| * Classification based on effluent monitor readings assumes that a release path
| |
| * If the effluent flow past an effluent monitor is known to have stopped, to the environment is established. If the effluent flow past an effluent monitor is indicating that the release path is isolated, the effluent monitor reading is known to have stopped due to actions to isolate the release path, then the no longer valid for classification purposes. effluent monitor reading is no longer valid for classification purposes.
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| : 1. Reading on ANY effluent radiation monitor greater than 2 times the (site- 1. Reading on ANY of the following effluent monitors > 2 times alarm specific effluent release controlling document) limits for 60 minutes or setpoint established by a current radioactive release discharge permit longer: for > 60 minutes.
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| (site-specific monitor list and threshold values corresponding to 2 times
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| * Radwaste Effluent Monitor 1/2-1799-01 the controlling document limits)
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| OR
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| * Discharge Permit specified monitor
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| : 2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for OR 60 minutes or longer. 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS
| |
| - Total Noble Gas Release Rate).
| |
| : 3. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times (site-specific effluent release OR controlling document limits) for 60 minutes or longer. 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
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| Page 4 of 66
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| NEI 99-01 Rev 6 Proposed EAL I Justification AG2 RG2 Initiating Condition -- GENERAL EMERGENCY Initiating Condition: F-- No Change E Difference 1 Deviation Spent fuel pool level cannot be restored to at least (site-specific Level 3 description)
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| Spent fuel pool level cannot be restored to at least (site-specific Level 3 for 60 minutes or longer.
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| description) for 60 minutes or longer. 1) EAL will be used in accordance with the discussion in Section 1.4, NRC Order Operating Mode Applicability: EA-12-051, it is recommended that this EAL be implemented when the enhanced Operating Mode Applicability: All spent fuel pool level instrumentation is available for use. The completion of the 1, 2, 3,4, 5, D enhanced SFP level indicators and need for the inclusion of this EAL is being Example Emergency Action Levels: Emergency Action Level (EAL): tracked in accordance with Exelon Generation Company, LLC's Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for NOTES: The Emergency Director should declare the General Emergency Note: The Emergency Director should declare the General Emergency promptly Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated promptly upon determining that 60 minutes has been exceeded, upon determining that the applicable time has been exceeded, or will likely October 25,2012.
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| or will likely be exceeded be exceeded.
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| : 1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for description) for 60 minutes or longer. 60 minutes or longer.
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| Page 5 of 66
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| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition - SITE AREA EMERGENCY Initiating Condition: No Change Difference Deviation Spent fuel pool level at (site-specific Level 3 description) Spent fuel pool level at (site-specific Level 3 description).
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| : 1) EAL will be used in accordance with the discussion in Section 1.4, NRC Order Operating Mode Applicability: All Operating Mode Applicability: EA-12-051, it is recommended that this EAL be implemented when the enhanced 1, 2, 3, 4, 5, D spent fuel pool level instrumentation is available for use. The completion of the Example Emergency Action Levels: enhanced SFP level indicators and need for the inclusion of this EAL is being Emergency Action Level (EAL): tracked in accordance with Exelon Generation Company, LLC's Initial Status
| |
| : 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). 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.
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| Page 6 of 66
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| NEI 99-01 Rev 6 1 Proposed EAL Justification AA2 RA2 Initiating Condition - ALERT Significant lowering of water level above, or damage to, irradiated fuel.
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| Initiating Condition: D No Change f Difference 1 Deviation Significant lowering of water level above, or damage to, irradiated fuel. 1) Listed site-specific monitors and Threshold values to ensure timely classification.
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| Operating Mode Applicability: All Operating Mode Applicability:
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| : 2) EAL #3 will be used in accordance with the discussion in Section 1.4, NRC 1, 2, 3,4, 5, D Order EA-1 2-051, it is recommended that this EAL be implemented when the Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): 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
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| : 1. Uncovery of irradiated fuel in the REFUELING PATHWAY. 1. Uncovery of irradiated fuel in the REFUELING PATHWAY. tracked in accordance with Exelon Generation Company, LLC's Initial Status OR Report to March 12, 2012 Commission Order Modifying Licenses with Regard for
| |
| : 2. Damage to irradiated fuel resulting in a release of radioactivity from the Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated fuel as indicated by ANY of the following radiation monitors: 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel October 25,2012.
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| as indicated by ANY Table RI Radiation Monitor reading >1000 mRem/hr.
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| (site-specific listing of radiation monitors, and the associated readings, OR setpoints and/or alarms)
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| : 3. Lowering of spent fuel pool level to (site-specific Level 2 value).
| |
| : 3. Lowering of spent fuel pool level to (site-specific Level 2 value).
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| 4.
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| Table RI Fuel Handling Incident Radiation Monitors
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| * 1(2) 1705-16A Fuel Pool Rad Monitor
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| * 1(2) 1705-168 Fuel Pool Rad Monitor Page 7 of 66
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| NEI 99-01 Rev 6 Proposed EAL [ Justification AU2 RU2 Initiating Condition: UNUSUAL EVENT Initiating Condition: E No Change E Difference 1
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| -- Deviation UNPLANNED loss of water level above irradiated fuel. UNPLANNED loss of water level above irradiated fuel 1) Listed site specific level indication and monitors to ensure timely Operating Mode Applicability: All Operating Mode Applicability: classification.
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| 1, 2, 3, 4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
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| : 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by indicated by ANY of the following: ANY of the following:
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| * Refueling Cavity water level < 282 in. (Upper Wide range (site-specific level indications). simulated signal).
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| OR AND
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| * Spent Fuel Pool water level <19 ft. above the fuel (<-4 ft.
| |
| indicated level).
| |
| OR
| |
| : b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
| |
| * Indication or report of a drop in water level in the REFUELING PATHWAY.
| |
| AND (site-specific list of area radiation monitors)
| |
| : b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table RI.
| |
| Table R1 Fuel Handling Incident Radiation Monitors
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| * 1(2) 1705-16A Fuel Pool Rad Mon
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| * 1(2) 1705-16B Fuel Pool Rad Mon Page 8 of 66
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| NEI 99-01 Rev 6 Proposed EAL Justification RA3 A3 E No Change 1 Difference 1 Deviation Initiating Condition:
| |
| Initiating Condition - ALERT Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or 1) Listed site specific plant rooms and areas with identified mode Radiation levels that impede access to equipment necessary for normal shutdown. applicability to ensure timely classification.
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| plant operations, cooldown or shutdown.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability: All 1, 2, 3, 4, 5,D Emergency Action Level (EAL):
| |
| Example Emergency Action Levels: (1 or 2)
| |
| Note:
| |
| Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no If the equipment in the room or area listed in Table R3 was already inoperable, or out of emergency classification is warranted service, before the event occurred, then no emergency classification is warranted.
| |
| : 1. Dose rate greater than 15 mR/hr in ANY of the following areas:
| |
| * Control Room 1. Dose rate> 15 mR/hr in ANY of the following Table R2 areas:
| |
| * Central Alarm Station Table R2 Areas Requiring Continuous Occupancy
| |
| * (other site-specific areas/rooms)
| |
| * Main Control Room (Unit 1 ARM Station #22)
| |
| : 2. An UNPLANNED event results in radiation levels that prevent
| |
| * Central Alarm Station - (by survey) or significantly impede access to any of the following plant rooms or areas: OR (site-specific list of plant rooms or areas with entry-related 2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to mode applicability identified) ANY of the following Table R3 plant rooms or areas:
| |
| Table R3 Areas with Entry Related Mode Applicability Entry Related Area Unit Mode Applicability Reactor Building
| |
| * First Floor North Wall 1
| |
| * Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Comer 2 Page 9 of 66
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| Justification Rev 66 Proposed EAL NEI 99-01 Rev NEI 99-01 Proposed EAL Justification SU3 RU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: H No Change E Difference I: Deviation Reactor coolant activity greater than Technical Specification allowable limits. Reactor coolant activity greater than Technical Specification allowable limits. 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| : 2) Listed this system category EAL in the radiological category EAL section to Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 maintain consistency with current and previous revisions of Exelon EALs. This will ensure a timely classification since the threshold values are more aligned with the Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| radiological category vice system category.
| |
| : 1. (Site-specific radiation monitor) reading greater than (site-specific value). 1. Offgas system radiation monitor HI-HI alarm.
| |
| OR OR
| |
| : 2. Sample analysis indicates that a reactor coolant activity value is greater than 2. Specific coolant activity > 4.0 uCi/gm Dose equivalent 1-131. -
| |
| an allowable limit specified in Technical Specifications.
| |
| Page 10 of 66
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| | |
| NEI 99-01 rev 6 Fission Product Barrier Matrix Fission____________
| |
| Prdc Mari _________________ Hot________
| |
| FGI Loss of any two barriers AND Loss or Potential Loss of third barrier. 1,2,3 FS1 Loss or Potential Loss of ANY two barriers. 1,2,3 FAI ANY Loss or ANY Potential Loss of either Fuel Clad or RCS 1,2,3 FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A. UNPLANNED rapid drop in A. Primary containment pressure
| |
| : 1. RCS Activity primary containment pressure greater than (site-specific value)
| |
| Primary A. (Site specific indications that A. Primary containment pressure follouin rime Containment reactor coolant activi is None greater than (site-specific None B. (site-specific explosive mixture)
| |
| Pressure / greater than 300 uCi/gm value) due to RCS leakage. OR exists inside primary containment Primary dose equivalent 1-131)exssiidprmycotnet Containment B. Primary containment pressure OR Conditions response not consistent with LOCA conditions. C. HCTL exceeded.
| |
| A. RPV water level cannot be A. RPV water level cannot be A. Primary containment flooding restored and maintained above restored and maintained above
| |
| : 2. RPV Water required. (site-specific RPV water level (site-specific RPV water level None None Level corresponding to top of active corresponding to the top of required.
| |
| active fuel) or cannot be fuel) or cannot be determined, determined.
| |
| A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR A. UNISOLABLE break in any of the A. UNISOLABLE primary system containment following: ( site-specific systems leakage thath results EIHRo in exceeding olwn:venting perprimary B. Intentional EOPs 3.RCS Leak with potential for high-energy line EITHER of the following:
| |
| Rate/ Primary None None breas) 1. Max Normal Operating OR None Containment OR Temperature C. UNISOLABLE primary system Isolation Failure B. Emergency RPV OR leakage that results in exceeding Depressurization 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 A. Primary Containment Radiation A. Primary Containment Radiation Containment Radiation Monitor reading None Monitor reading greater than None None Monitor reading greater than (site-Raiation greater than (site-specific (site-specific value).
| |
| Radiation specific value).
| |
| value).
| |
| : 5. Emergency A. Any Condition in the opinion A.Any Condition in the opinion of A. ANY Condition in the opinion of A. Any Condition in the opinion of the A.Any Condition in the opinion of the A. Any Condition in the opinion of the Director of the Emergency Director the Emergency Director that the Emergency Director that Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates that indicates Loss of the indicates Potential Loss of the indicates Loss of the RCS Potential Loss of the RCS Barrier. Loss of the Containment Barrier. Potential Loss of the Containment Judgment Fuel Clad Barrier. Fuel Clad Barrier. Barrier. P I Barrier.
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| Page II of 66
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| | |
| PrnrnnserlPinnirin Prorlirot flurries Matrin Fl Loss of any two barriers AND Loss or Potential Loss of third barrier. PSi Loss or Potential Loss of ANY two barriers. EFAl ANY Loss or ANY Potential Loss of either Fuel Clad or RCS i FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity > 300 uCi/gm Dose None Nuns
| |
| : 1. RCS Activity Equivalent 1-131. NnNoeNone None None
| |
| : 2. RPV water level cannot be restored and 1. RPV water level cannot be restored and 21 RPV Water 1. Plant conditions indicate Primary maintained > -142 inches (TAF) maintained >-142 inches (TAF) Plant conditions indicate Primary Containment 2Rver Containment flooding is required. OR OR None None flooding is required.
| |
| Level 3. RPV water level cannot be determined. 2. RPV water level cannot be determined.
| |
| : 3. Primary Containment pressure > 56 pslg.
| |
| OR
| |
| : 1. UNPLANNED rapid drop in Drywell 4. a. Drywellor torus hydrogen concentration
| |
| : 1. Drywallpressure '2.5 psig. pressure followingprimary containment > 6%.
| |
| : 3. Primary AND pressure rise. AND Containment None None 2. Drywellpressure rise is due to RCS None OR b. Drywellor tolrs oxygen concentration Pressure/Conditions leakage 2. Drywellpressure response not consistent _0%.
| |
| with LOCAconditions. OR
| |
| : 5. Heat Capacity Limit(OGA200. Figure M) exceeded.
| |
| : 3. UNISOLABLE primary system leakage that
| |
| : 1. UNISOLABLE MainSteam Line (MSL), results in EITHER of the following:
| |
| HPCI. Feedwater. RWCU or RCIC line a. Secondary Containment area break. temperature n OGA 300 Maximum 4.RCS Leak Rate None None OR Normal operating levels.
| |
| : 2. Emergency RPV Depressurization is OR required. b.Secondary Containment area radiation level > QGA300 Maximum Normeal operating level.
| |
| 5.Primary Drywellradiation monitor reading Drywellradiation monitor reading Drywellradiation monitor reading Containment > 6.65 Ev02 R/hr. None >10OR/thr (>t.00 E+02 R/hr). None None > 1.55 E+03 Rthr Radiation
| |
| : 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 ofthe following:
| |
| : a. Secondary Containment area temperature > QGA300, Maximum Safe operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > OGA 300, Maximum Safe operating levels
| |
| : 7. Emergen A.AnyCondition in the opinion of the A. Any Condition in the opinion of the A.ANYCondition in the opinion of the A.AnyCondition in the opinion of the A.Any Condition in the opinion of the A.Any Condition in the opinion of the 7.Eergecy 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 Barner. of the Containment Barrier.
| |
| Page 12 of66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification FCl FC1 Category: Fuel Clad Barrier Category: Fuel Clad Barrier E No Change -- 1 Difference 1 Deviation RCS Activity RCS Activity 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. (Site specific indications that reactor coolant activity is greater than 300 uCi/gm Coolant activity > 300 uCi/gm Dose Equivalent 1-131.
| |
| dose equivalent 1-131)
| |
| Page 13 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification FC2 FC2 I-- No Change 1 Difference 1 Deviation Category: Fuel Clad Barrier Category: Fuel Clad Barrier RPV Water Level RCS Activity 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary containment flooding required. 1. Plant conditions indicate Primary Containment flooding is required.
| |
| Potential Loss Potential Loss A. RPV water level cannot be restored and maintained above (site-specific RPV 2. RPV water level cannot be restored and maintained > -142 inches (TAF) water level corresponding to top of active fuel) or cannot be determined.
| |
| OR
| |
| : 3. RPV water level cannot be determined.
| |
| Page 14 of66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification FC4 FC5 Category: Fuel Clad Barrier Category: Fuel Clad Barrier No Change 1 Difference 1 Deviation Primary Containment Radiation Primary Containment Radiation 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywall radiation monitor reading > 6.65 E+02 R/hr.
| |
| value).
| |
| Page 15 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification FC6 FC7 Category: Fuel Clad Barrier Category: Fuel Clad Barrier 1-- No Change D Difference 1 Deviation Emergency Director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. Any Condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier. the Fuel Clad Barrier.
| |
| Potential Loss Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier. Loss of the Fuel Clad Barrier.
| |
| Page 16 of 66
| |
| | |
| NEI 99-01 Rev 6 j Proposed EAL Justification RCI RC3 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change D Difference 1: Deviation Primary Containment Pressure Primary Containment Pressure/Conditions 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary containment pressure greater than (site-specific value) due to RCS 1. Drywell pressure >2.5 psig.
| |
| leakage. AND
| |
| : 2. Drywell pressure rise is due to RCS leakage.
| |
| Page 17 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification RC2 RC2 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change L Difference L 1 Deviation RPV Water Level RPV Water Level 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. RPV water level cannot be restored and maintained above (site-specific RPV 1. RPV water level cannot be restored and maintained > -142 inches (TAF).
| |
| level corresponding to the top of active fuel) or cannot be determined. OR
| |
| : 2. RPV water level cannot be determined.
| |
| Page 18 of66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification RC3 RC4 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier W No Change 1 Difference 1 Deviation RCS Leak Rate RCS Leak Rate 1) Listed site-specific systems and threshold values to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. UNISOLABLE break in any of the following: ( site-specific systems with potential 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater,, RWCU or RCIC for high-energy line breas) line break.
| |
| OR OR
| |
| : 2. Emergency RPV Depressurization is required.
| |
| B. Emergency RPV Depressurization Potential Loss Potential Loss A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following: 3. UNISOLABLE primary system leakage that results in EITHER of the following:
| |
| : a. Secondary Containment area temperature > QGA 300 Maximum Normal
| |
| : 1. Max Normal Operating Temperature. operating levels.
| |
| OR OR
| |
| : b. Secondary Containment area radiation level > QGA 300 Maximum Normal
| |
| : 2. Max Normal Operating Area Radiation Level.
| |
| operating levels.
| |
| Page 19 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification RC4 RC5 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change 1 Difference 1: Deviation Primary Containment Radiation Primary Containment Radiation 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 100 R/hr (>1.00E+02 R/hr).
| |
| value).
| |
| Page 20 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification RC6 RC7 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change D Difference 1 Deviation Emergency Director Judgment Emergency director Judgment Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. ANY Condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier. the RCS Barrier.
| |
| Potential Loss Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier. 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.
| |
| Page 21 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification CT1 CT3 Category: Containment Barrier Category: Containment Barrier D No Change E Difference I: Deviation Primary Containment Conditions Primary Containment Pressure/Conditions 1) Listed site-specific threshold values to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| : 2) The words "and rising" were added to account for the momentary spike in Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 pressure where pressure is now lowering, the risk of a potential loss of Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: containment is no longer present, this wording is also consistent with present EAL wording.
| |
| Loss Loss C. UNPLANNED rapid drop in primary containment pressure following primary 1. UNPLANNED rapid drop in Drywell pressure following Drywall pressure rise.
| |
| containment pressure rise OR OR 2. Drywell pressure response not consistent with LOCA conditions.
| |
| B. Primary containment pressure response not consistent with LOCA conditions. Potential Loss Potentiat Loss
| |
| : 3. Drywell pressure > 56 psig and rising.
| |
| D. Primary containment pressure greater than (site-specific value)
| |
| OR OR 4. a. Drywell or torus Hydrogen concentration > 6%.
| |
| E. (site-specific explosive mixture) exists inside primary containment AND OR b. Drywell or torus Oxygen concentration > 5%.
| |
| OR
| |
| : 3. HCTL exceeded. 5. Heat Capacity Limit (QGA 200, Fig. M) exceeded.
| |
| Page 22 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CT2 CT2 Category: Containment Barrier Category: Containment Barrier Ejj No Change jjj Difference 1-1 Deviation RPV Water Level RPV Water Level Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Potential Loss Potential Loss A. Primary containment flooding required.
| |
| Plant conditions indicate Primary Containment flooding is required.
| |
| Page 23 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification CT3 CT6 Category: Containment Barrier Category: Containment Barrier 1fl No Change E Difference r-- Deviation Primary Containment Isolation Failure Primary Containment Isolation Failure 1) Listed site-specific threshold values to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. UNISOLABLE direct downstream pathway to the environment exists after 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal primary containment isolation signal.
| |
| OR OR
| |
| : 2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to B. Intentional primary containment venting per EOPs accident conditions.
| |
| OR OR C. UNISOLABLE primary system leakage that results in exceeding EITHER of the 3. UNISOLABLE primary system leakage that results in EITHER of the following:
| |
| following: a. Secondary Containment area temperature > QGA 300 Maximum Safe operating levels.
| |
| : 1. Max Safe Operating Temperature. OR OR b. Secondary Containment area radiation level > QGA 300 Maximum Safe
| |
| : 2. Max Safe Operating Area Radiation Level. operating levels.
| |
| Page 24 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CT4 CT5 Category: Containment Barrier Category: Containment Barrier No Change 1 Difference 1 Deviation Primary Containment Radiation Primary Containment Radiation 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Potential Loss Potential Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 1.55 E+03 R/h.
| |
| value).
| |
| Page 25 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CT6 CT7 Category: Containment Barrier No Change Difference Deviation Category: Containment Barrier Emergency director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. Any Condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier. the Containment Barrier.
| |
| Potential Loss Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier. 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.
| |
| Page 26 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SG1 MG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition:
| |
| I-* No Change EIDifference -- 1 Deviation Prolonged loss of all offsite and all onsite AC power to emergency buses. Prolonged loss of all offsite and all onsite AC power to emergency buses.
| |
| : 1) Listed site specific equipment, site specific time based on station blackout Operating Mode Applicability: Operating Mode Applicability: coping analysis, and site specific indication to ensure timely classification.
| |
| 1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the General Emergency promptly upon determining that (site-specific hours) has been exceeded, or will Note:
| |
| likely be exceeded.
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| : 1. a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).
| |
| 1 Loss of ALL offsite AC power to unit ECCS busses.
| |
| AND AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel
| |
| : b. EITHER of the following: generators to supply power to unit ECCS busses.
| |
| AND
| |
| * Restoration of at least one emergency bus in less than (site-specific hours) is not likely. 3. EITHER of the following:
| |
| : a. Restoration of at least one unit ECCS bus in < I hour is not likely.
| |
| * (Site-specific indication of an inability to adequately OR remove heat from the core)
| |
| : b. RPV water level cannot be restored and maintained > -166 Inches.
| |
| Page 27 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification SS1 MS1 Initiating Condition: SITE AREA EMERGENCY Initiating Condition:
| |
| Loss of all offsite and onsite AC power to emergency buses for 15 minutes or M No Change D Difference FIDeviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.
| |
| longer.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| 1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Site Area Emergency Note:
| |
| promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) exceeded.
| |
| for 15 minutes or longer.
| |
| : 1. Loss of ALL offsite AC Power to unit ECCS buses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Page 28 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification SA1I MA1I Initiating Condition: ALERT Initiating Condition:
| |
| -<I No Change [ ]Difference F Deviation Loss of all but one AC power source to emergency buses for 15 minutes or longer. Loss of all but one AC power source to emergency buses for 15 minutes or longer.
| |
| : 1) Listed site specific equipment to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:
| |
| : 1. a. AC power capability to (site-specific emergency buses) is reduced to a The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be single power source for 15 minutes or longer. exceeded.
| |
| AND
| |
| : 1. AC power capability to unit ECCS buses reduced to only one of the following
| |
| : b. Any additional single power source failure will result in loss of all AC power sources for > 15 minutes.
| |
| power to SAFETY SYSTEMS.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit Auxiliary Transformer TR-11 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers AND
| |
| : 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.
| |
| Page 29 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification MUl Initiating Condition: UNUSUAL EVENT Initiating Condition:
| |
| FI No Change EIDifference F- Deviation Loss of all offsite AC power capability to emergency buses for 15 minutes or Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. longer. 1) Listed site specific equipment to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 exceeded.
| |
| minutes or longer Loss of ALL offsite AC power capability to unit ECCS buses for > 15 minutes.
| |
| Page 30 of 66
| |
| | |
| NEI 99-01 Rev 6 J Proposed EAL Justification Initiating Condition: GENERAL EMERGENCY Loss of all AC and Vital DC power sources for 15 minutes or longer.
| |
| SG8 Initiating Condition:
| |
| MG2 D No Change M Difference I Deviation Loss of all AC and Vital DC power sources for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.
| |
| Operating Mode Applicability: 2) in EAL #3 removed the word "indicated" this will allow for an indication problem Operating Mode Applicability: to not cause confusion on the need to declare.
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded.
| |
| : 1. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be buses) for 15 minutes or longer. exceeded.
| |
| AND
| |
| : 1. Loss of ALL offsite AC power to unit ECCS busses.
| |
| Indicated voltage is less than (site-specific bus voltage value) on ALL AND (site-specific vital DC buses) for 15 minutes or longer.
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.
| |
| AND
| |
| : 3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.
| |
| AND
| |
| : 4. ALL AC and Vital DC power sources have been lost for > 15 minutes.
| |
| Page 31 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification ss8 MS2 Initiating Condition: SITE AREA EMERGENCY Loss of all Vital DC power for 15 minutes or longer.
| |
| Initiating Condition: E] No Change H- Difference 1- Deviation Loss of all Vital DC power for 15 minutes or longer.
| |
| : 1) Listed site specific equipment and value to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability: 2) Removed the word "indicated" this will allow for an indication problem to not Power Operation, Startup, Hot Standby, Hot Shutdown cause confusion on the need to declare.
| |
| 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:
| |
| Indicated voltage is less than (site-specific bus voltage value) on ALL Vital DC
| |
| * The Emergency Director should declare the event promptly upon buses for 15 minutes or longer. determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| Voltage is < 105 VDC on 125 VDC battery buses #1 and #2 for >15 minutes..
| |
| Page 32 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SS5 MS3 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: D No Change FIDifference F Deviation Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPV water level [BWR]) or RCS heat removal. Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. 1) Listed site specific indications to ensure timely classification.
| |
| Operating Mode Applicability: 2) Mode 2 included in operating mode applicability as per developer notes.
| |
| Operating Mode Applicability:
| |
| Power Operation 1,2 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| : 1. a. An automatic (trip [PWRI / scram [BWR]) did not shutdown the reactor.
| |
| : 1. Automatic scram did not shutdown the reactor as indicated by Reactor Power AND > 5%.
| |
| AND
| |
| : b. All manual actions to shutdown the reactor have been unsuccessful. 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
| |
| AND AND
| |
| : c. EITHER of the following conditions exist: 3. EITHER of the following conditions exist:
| |
| : 1. (Site-specific indication of an inability to adequately " RPV water level cannot be restored and maintained > -166 inches.
| |
| remove heat from the core) OR
| |
| " Heat Capacity Limit (CGA 200, Figure M) exceeded.
| |
| OR
| |
| : 2. (Site-specific indication of an inability to adequately remove heat from the RCS)
| |
| Page 33 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SA5 MA3 Initiating Condition: ALERT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and Initiating Condition: F No Change M Difference [-] Deviation Automatic or manual scram fails to shutdown the reactor, and subsequent manual 1) Listed site specific indications to ensure timely classification.
| |
| subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. actions taken at the reactor control consoles are not successful in shutting down the reactor. 2) Mode 2 included in operating mode applicability as per developer notes.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation 1,2 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually Note:
| |
| driving in control rods or implementation of boron injection strategies.
| |
| A manual action is any operator action, or set of actions, which causes the
| |
| : 1. a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor.
| |
| manually driving in control rods or implementation of boron injection strategies.
| |
| AND
| |
| : 1. Automatic or manual scram did not shutdown the reactor as indicated by
| |
| : b. Manual action taken at the reactor control consoles are not Reactor Power > 5%.
| |
| successful in shutting down the reactor.
| |
| AND
| |
| : 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.
| |
| Page 34 of 66
| |
| | |
| Justification Proposed EAL NEI Rev 66 99-01 Rev NEt 99-01 Proposed EAL Justification SU5 MU3 Initiating Condition: UNUSUAL EVENT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor.
| |
| Initiating Condition: F No Change W Difference FI Deviation Automatic or manual scram fails to shutdown the reactor.
| |
| : 1) Listed site specific indications to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| : 2) Mode 2 included in operating mode applicability as per developer notes.
| |
| Power Operation 1,2 Example Emergency Action Levels: (1 or 2)
| |
| Emergency Action Level (EAL):
| |
| Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note:
| |
| A manual action is any operator action, or set of actions, which causes the
| |
| : 1. a. An automatic (trip [PWRJ / scram [BWR]) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor. manually driving in control rods or implementation of boron injection strategies.
| |
| AND
| |
| : a. Automatic scram did not shutdown the reactor as indicated by Reactor
| |
| : b. A subsequent manual action taken at the reactor control consoles is Power > 5%.
| |
| successful in shutting down the reactor.
| |
| AND
| |
| : b. Subsequent manual / ARI action taken at the Reactor Console is
| |
| : 2. a. A manual scram ([PWR] / scram [BWR]) did not shutdown the successful in shutting down the reactor.
| |
| reactor.
| |
| OR AND 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power > 5%.
| |
| : b. EITHER of the following: AND
| |
| : b. EITHER of the following:
| |
| : 1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. 1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.
| |
| OR OR
| |
| : 2. Subsequent automatic scram / ARI is successful in shutting down
| |
| : 2. A subsequent automatic (trip [PWR] / scram [BWRJ) is successful the reactor.
| |
| in shutting down the reactor.
| |
| Page 35 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SA2 MA4 Initiating Condition: ALERT Initiating Condition: F- No Change DIDifference F Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Note: The Emergency Director should declare the Alert promptly upon Emergency Action Level (EAL):
| |
| determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded.
| |
| The Emergency Director should declare the event promptly upon
| |
| : 1. a. An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.
| |
| [see table below]
| |
| : 1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter
| |
| [ BWR parameter list] [PWR parameter list] from within the Control Room for >15 minutes.
| |
| Reactor Power Reactor Power Table M1 Control Room Parameters RPV Level RCS Level
| |
| " Reactor Power
| |
| " RPV Water Level RPV Pressure RCS Pressure " RPV Pressure
| |
| " Drywell Pressure Primary Containment Pressure In Core/Core Exit Temperature " Torus Level
| |
| " Torus Temperature Levels in at least (site specific Suppression Pool Level number) steam generators AND Suppression Pool Temperature Steam Generator Auxiliary or 2. ANY Table M2 transient in progress.
| |
| Emergency Feed Water Flow AND Table M2 Significant Transients
| |
| : b. Any of the following transient events in progress. " Turbine Trip
| |
| " Automatic or Manual runback greater than 25% thermal reactor power " Reactor Scram
| |
| " Electrical load rejection greater than 25% full electrical load " ECCS Activation
| |
| " Reactor Scram [BWR] / trip [PWR]
| |
| " ECCS (SI) actuation " Recirc. Runback > 25% Reactor Power Change
| |
| " Thermal power oscillations greater than 10% [BWR] " Thermal Power oscillations > 10% Reactor Power Change Page 36 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SU2 MU4 Initiating Condition: UNUSUAL EVENT UNPLANNED loss of Control Room indications for 15 minutes or longer.
| |
| Initiating Condition: F- No Change IDifference F Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):
| |
| determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded.
| |
| The Emergency Director should declare the event promptly upon An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.
| |
| [see table below] UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for > 15 minutes.
| |
| [ BWR parameter list] [ PWR parameter list]
| |
| Table MI Control Room Parameters Reactor Power Reactor Power
| |
| " Reactor Power RPV Level RCS Level " RPV Water Level
| |
| " RPV Pressure
| |
| " Drywell Pressure RPV Pressure RCS Pressure
| |
| " Torus Level
| |
| " Torus Temperature Primary Containment Pressure InCore(Core Exit Temperature Levels in at least (site specific Suppression Pool Level number) steam generators Suppression nPool Temperature PEmergency Steam Generator Auxiliary Feed Water or Flow Page 37 of66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SA9 MA5 Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM needed for the current operating Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM required for the current operating FD1No Change
| |
| * Difference F Deviation
| |
| : 1) No additional site specific hazard noted mode. mode.
| |
| : 2) Changed the word "needed" to "required" in the IC and "required by Technical Operating Mode Applicability: Operating Mode Applicability: Specification" in the EAL to be consistent with terminology used by operators and minimize confusion.
| |
| 1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Level (EAL): 3) Added additional note to easily direct the operator to lesser ICs Note:
| |
| Example Emergency Action Levels:
| |
| If it is determined that the conditions of MA5 are not met then assess the
| |
| : 1. a. The occurrence of ANY of the following hazardous events: event via HU3, HU4, or HU6.
| |
| * Seismic event (earthquake)
| |
| The occurrence of ANY of the following hazardous events:
| |
| * Internal or extemal flooding event
| |
| * Seismic event (earthquake)
| |
| * High winds or tornado strike
| |
| * FIRE " Internal or external flooding event
| |
| * EXPLOSION
| |
| * High winds or tornado strike
| |
| * (site-specific hazards)
| |
| * FIRE
| |
| * Other events with similar hazard characteristics as determined by the Shift Manager
| |
| * EXPLOSION
| |
| * Other events with similar hazard characteristics as determined AND by the Shift Manager AND
| |
| : b. EITHER of the following:
| |
| : 2. EITHER of the following:
| |
| : 1. Event damage has caused indications of degraded performance a. Event damage has caused indications of degraded in at least one train of a SAFETY SYSTEM needed for the performance in at least one train of a SAFETY SYSTEM current operating mode. required by Technical Specifications for the current operating mode.
| |
| OR OR
| |
| : b. The event has caused VISIBLE DAMAGE to a SAFETY
| |
| : 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM SYSTEM component or structure required by Technical component or structure needed for the current operating mode.
| |
| Specifications for the current operating mode.
| |
| Page 38 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification SU4 MU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change FIDifference FIDeviation RCS leakage for 15 minutes or longer.
| |
| RCS leakage for 15 minutes or longer. 1) Listed site specific values to ensure timely classification.
| |
| Operating Mode Applicability: 2) Changed wording from containment to Drywall for clarity to better define the Operating Mode Applicability: primary containment structure.
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 3) In EAL #1 and 2 added "into the Drywell" to differentiate between EAL #1/2 and
| |
| #3. Without this wording would have been in EAL #1 or #2 concurrent with #3. With Example Emergency Action Levels: (1 or 2 or 3) the added wording each EAL can be called separately.
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Note:
| |
| determining that 15 minutes has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be
| |
| : 1. RCS unidentified or pressure boundary leakage greater than (site-specific exceeded.
| |
| value) for 15 minutes or longer.
| |
| : 1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for
| |
| : 2. RCS identified leakage greater than (site-specific value) for 15 minutes or > 15 minutes.
| |
| longer OR
| |
| : 2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.
| |
| : 3. Leakage from the RCS to a location outside containment greater than 25 OR gpm for 15 minutes or longer
| |
| : 3. Leakage from the RCS to a location outside the Drywell >25 gpm for
| |
| > 15 minutes.
| |
| Page 39 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification SU6 Initiating Condition: UNUSUAL EVENT Loss of all onsite or offsite communications capabilities Initiating Condition:
| |
| MU7 W No Change E Difference F Deviation Loss of all onsite or offsite communication capabilities. 1) Listed site specific communication methods to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1.2,3 Example Emergency Action Levels: (1 or 2 or 3)
| |
| Emergency Action Level (EAL):
| |
| : 1. Loss of ALL of the following onsite communication methods:
| |
| (site-specific list of communications method 1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine operations.
| |
| : 2. Loss of ALL of the following ORO communications s) methods: OR
| |
| : 2. Loss of ALL Table M3 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.
| |
| OR
| |
| : 3. Loss of ALL of the following NRC communications methods:
| |
| : 3. Loss of ALL Table M3 NRC communication capability affecting the ability to perform NRC notifications.
| |
| (site-specific list of communications methods)
| |
| Table M3 Communications Capabil'y -
| |
| System Onsite Offsite NRC Plant Radio X Plant Page X All telephone Lines (Commercial and X X X microwave)
| |
| ENS X X HPN X X Satellite Phones X X Page 40 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification CA2 CAl Initiating Condition: ALERT Initiating Condition: E No Change 1 Difference I: Deviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.
| |
| longer.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| 4,5, D Cold Shutdown, Refueling, Defueled Emergency Action Level (EAL):
| |
| Example Emergency Action Levels:
| |
| Note:
| |
| Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes time has been exceeded, or will likely be The Emergency Director should declare the event promptly upon determining exceeded. that the applicable time has been exceeded, or will likely be exceeded.
| |
| Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 minutes or longer. 1. Loss of ALL offsite AC power to unit ECCS buses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Page 41 of 66
| |
| | |
| Justification Proposed EAL NEI 99-01 Rev NEI Rev 66 Proposed EAL Justification CU2 Cut Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change 1:1 Difference 1 Deviation Loss of all but one AC power source to emergency buses for 15 minutes or longer. Loss of all but one AC power source to emergency buses for 15 minutes or 1) Listed site specific equipment to ensure timely classification.
| |
| longer.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Cold Shutdown, Refueling, Defueled 4,5, D Example Emergency Action Levels:
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):
| |
| determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:
| |
| : 1. a. AC power capability to (site-specific emergency buses) is reduced to a The Emergency Director should declare the event promptly upon single power source for 15 minutes or longer. determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| AND
| |
| : 1. AC power capability to unit ECCS buses reduced to only one of the
| |
| : b. Any additional single power source failure will result in loss of all AC following power sources for > 15 minutes.
| |
| power to SAFETY SYSTEMS.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit auxiliary transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers AND
| |
| : 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.
| |
| Page 42 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification Initiating Condition - ALERT CA6 Initiating Condition:
| |
| CA2 D-1 No Change E Difference -- Deviation Hazardous event affecting SAFETY SYSTEM needed for the current operating Hazardous event affecting SAFETY SYSTEM required for the current operating 1) No additional site specific hazards noted mode. mode.
| |
| Operating Mode Applicability: Operating Mode Applicability: 2) Changed the word "needed" to "required" in the IC and "required by Technical Specification" in the EAL to be consistent with terminology used by Cold Shutdown, Refueling 4,5 operators and minimize confusion.
| |
| Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| : 1. a. The occurrence of ANY of the following hazardous events: Note: 3) Added additional note to easily direct the operator to lesser ICs
| |
| " Seismic event (earthquake) If it is determined that the conditions of CA2 are not met then assess the
| |
| " Internal or external flooding event event via HU3, HU4, or HU6.
| |
| * High winds or tornado strike The occurrence of ANY of the following hazardous events:
| |
| " FIRE
| |
| * Seismic event (earthquake) i EXPLOSION
| |
| * Internal or external flooding event
| |
| * (site-specific hazards) " High winds or tornado strike
| |
| " Other events with similar hazard characteristics as
| |
| * FIRE determined by the Shift Manager " EXPLOSION AND
| |
| * Other events with similar hazard characteristics as
| |
| : b. EITHER of the following: determined by the Shift Manager
| |
| : 1. Event damage has caused indications of degraded AND performance in at least one train of a SAFETY SYSTEM 2. EITHER of the following:
| |
| needed for the current operating mode.
| |
| : a. Event damage has caused indications of degraded OR performance in at least one train of a SAFETY SYSTEM
| |
| : 2. The event has caused VISIBLE DAMAGE to a SAFETY required by Technical Specifications for the current SYSTEM component or structure needed for the current operating mode.
| |
| 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.
| |
| Page 43 of 66
| |
| | |
| Justification Proposed EAL NEI 99-01 Rev 6 Proposed EAL Justification CU4 CU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change E Difference [: Deviation Loss of Vital DC power for 15 minutes or longer. Loss of Vital DC power for 15 minutes or longer. 1) Listed site specific voltage and equipment to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| : 2) Removed the word "indicated" this will allow for an indication problem to not Cold Shutdown, Refueling 4,5 cause confusion on the need to declare.
| |
| Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Note:
| |
| determining that 15 minutes time has been exceeded, or will likely be The Emergency Director should declare the event promptly upon exceeded.
| |
| determining that the applicable time has been exceeded, or will likely be Indicated voltage is less than (site-specific bus voltage value) on required Vital DC exceeded.
| |
| buses for 15 minutes or longer.
| |
| Voltage is < 105 VDC on required 125 VDC battery buses #1 and #2 for > 15 minutes.
| |
| Page 44 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT Loss of all onsite or offsite communications capabilities CU5 Initiating Condition:
| |
| CU4 W No Change ED ifference -- 1 Deviation Loss of all onsite or offsite communication capabilities. 1) Listed site specific communications methods to ensure timely classification Operating Mode Applicability: Operating Mode Applicability:
| |
| Cold Shutdown, Refueling, Defuled 4,5, D Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL):
| |
| : 1. Loss of ALL of the following onsite communication methods: 1. Loss of ALL Table C1 Onsite communications capability affecting the (site-specific list of communications method ability to perform routine operations.
| |
| OR
| |
| : 2. Loss of ALL of the following ORO communications s) methods:
| |
| : 2. Loss of ALL Table C1 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.
| |
| : 3. Loss of ALL of the following NRC communications methods:
| |
| OR (site-specific list of communications methods)
| |
| : 3. Loss of ALL Table C1 NRC communication capability affecting the ability to perform NRC notifications.
| |
| Table Cl Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X All telephone Lines (Commercial and X X X microwave)
| |
| ENS X X HPN X X Satellite Phones X X Page 45 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification CA3 CA5 D NoChange W Difference D eviation Initiating Condition: ALERT Initiating Condition:
| |
| Inability to maintain the plant in cold shutdown. Inability to maintain plant in cold shutdown.
| |
| Operating Mode Applicability: Operating Mode Applicability: 1) Listed site specific Technical Specification cold shutdown temperature limit to ensure timely classification.
| |
| Cold Shutdown, Refueling 4,5
| |
| : 2) Listed site specific pressure reading to enhance timely classification.
| |
| Example Emergency Action Levels: (1 or 2) Emergency Action Levels (EAL):
| |
| Note: The Emergency Director should declare the Alert promptly upon Note: The Emergency Director should declare the event promptly upon 3) Added wording relating the temp and press rise to a loss of decay heat removal determining that the applicable has been exceeded, or will likely be determining that the applicable time has been exceeded, or will likely be capability as per the developer notes for PWR's exceeded. exceeded.
| |
| : 4) Added the following note, taken from the basis section of the EAL, to bring it to
| |
| : 1. UNPLANNED increase in RCS temperature to greater than (site-specific A momentary UNPLANNED excursion above the Technical Specification the attention of the SM/ED when using the "procedure matrix" (11x17 quick Technical Specification cold shutdown temperature limit) for greater than cold shutdown temperature limit when heat removal function is available reference control room document) "Amomentary UNPLANNED excursion above the duration specified in the following table. does not warrant classification. the Technical Specification cold shutdown temperature limit when heat removal 0
| |
| : 2. UNPLANNED RCS pressure increase greater than (site-specific pressure 1. UNPLANNED rise in RCS temperature > 212 F for > Table C2 duration. function is available does not warrant classification."
| |
| reading). (This EAL does not apply during water-solid plant conditions.
| |
| [PWR])
| |
| OR
| |
| : 2. UNPLANNED RCS pressure rise > 10 psig as a result of temperature Table: RCS Heat-up Duration Thresholds rise.
| |
| RCS Status Containment Closure Heat-up Duration Status Table C2 - RCS Heat-up Duration Thresholds Intact (but not RCS Status Containment Closure Heat-up Duration RCS Reduced Not Applicable 60 minutes* Status Inventory (PWR]) Intact Not Applicable 60 minutes*
| |
| Not Intact (or at Established 20 minutes*
| |
| reduced inventory
| |
| [PWR]) Not Established 0 minutes Established 20 minutes*
| |
| * If an RCS heat removal system is in operation within this time frame and Not Intact RCS temperature is being reduced, the EAL is not applicable.
| |
| 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 aonlicable.
| |
| Page 46 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CU3 CU5 Initiating Condition: UNUSUAL EVENT Initiating Condition: H No Change W Difference 1 Deviation UNPLANNED increase in RCS temperature. UNPLANNED rise in RCS temperature. 1) Listed site specific Technical Specification cold shutdown temperature limit Operating Mode Applicability: Operating Mode Applicability: to ensure timely classification.
| |
| Cold Shutdown, Refueling 4,5 2) Changed the word increase to rise in the initiating condition to be consistent Example Emergency Action Levels: (1 or 2) with operations language and training.
| |
| Emergency Action Levels (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly Note: The Emergency Director should declare the event promptly upon 3) Added the following note, taken from the basis section of the EAL, to bring it upon determining that 15 minutes time has been exceeded, or will likely determining that the applicable time has been exceeded, or will likely to the attention of the SM/ED when using the "procedure matrix" (1lx17 quick be exceeded. be exceeded. reference control room document) "A momentary UNPLANNED excursion
| |
| : 1. UNPLANNED increase in RCS temperature to greater than (site-specific above the Technical Specification cold shutdown temperature limit when heat A momentary UNPLANNED excursion above the Technical removal function is available does not warrant classification."
| |
| Technical Specification cold shutdown temperature limit). Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.
| |
| : 2. Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV 0
| |
| [BWRJ) level indication for 15 minutes or longer. 1. UNPLANNED rise in RCS temperature > 212 F.
| |
| OR
| |
| : 2. Loss of the following for > 15 minutes.
| |
| * ALL RCS temperature indications AND
| |
| * ALL RPV level indications Page 47 of66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification CGI CG6 Initiating Condition: GENERAL EMERGENCY Initiating Condition: H No Change E Difference 1-- Deviation Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting fuel clad Loss of reactor vessel / RCS inventory affecting fuel clad integrity with integrity with containment challenged. containment challenged. 1) Listed site specific levels, radiation monitors, and sumps and tanks to ensure Operating Mode Applicability: Operating Mode Applicability: timely classification.
| |
| Cold Shutdown, Refueling 4,5 2) Listed Explosive mixture in the Containment Challenge Table to ensure timely classification.
| |
| Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the General Emergency Note:
| |
| promptly upon determining that 30 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be
| |
| : 1. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) vessel level less than (site- exceeded.
| |
| specific level) for 30 minutes or longer.
| |
| : 1. a. RPV water level < -142 inches (TAF) for > 30 minutes.
| |
| AND AND
| |
| : b. ANY indication from the Containment Challenge Table b. Any Containment Challenge Indication (Table C4)
| |
| : 2. a.. (Reactor vessel/RCS [PWR] or RPV [BWR]) vessel level cannot be OR monitored for 30 minutes or longer. 2. a. RPV water level cannot be monitored for > 30 minutes.
| |
| AND AND
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| * Table C3 indications of a sufficient magnitude to indicate core
| |
| * (Site-specific radiation monitor) reading greater than (site-specific value) uncovery.
| |
| * Erratic source range monitor indication [PWR] OR
| |
| * UNPLANNED increase in (site-specific sump and/or tank levels) of
| |
| * Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.
| |
| sufficient magnitude to indicate core uncovery AND
| |
| * (Other site-specific indications) c. ANY Containment Challenge Indication (Table C4)
| |
| AND Table C3 Indications of RCS Leakage
| |
| : c. ANY indication from the Containment Challenge Table).
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| Table: Containment Challenge Table
| |
| * UNPLANNED Torus level rise*
| |
| * CONTAINMENT CLOSURE not established*
| |
| * (Explosive mixture) exists inside containment
| |
| * UNPLANNED vessel make up rate rise
| |
| * UNPLANNED increase in containment pressure
| |
| * Observation of leakage or inventory loss
| |
| * Secondary containment radiation monitor reading above (site-specific value) [BWR] *Rise in level is attributed to a loss of RPV inventory.
| |
| if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not reouired. Table C4 Containment Challenge Indications
| |
| * Primary Containment Hydrogen Concentration > 6% and Oxygen
| |
| > 5%
| |
| * UNPLANNED rise in containment pressure
| |
| * CONTAINMENT CLOSURE not established*
| |
| * ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.
| |
| if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.
| |
| __
| |
| __ _ __ __ _ _
| |
| __ __ _ _I__ _ _ _ . 'I Page 48 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification Initiating Condition: SITE AREA EMERGENCY CS1 Initiating Condition:
| |
| CS6 D No Change E Difference 1: Deviation
| |
| : 1) Listed site specific values for level, radiation monitors, and sumps and tanks Loss of (reactor vesselIRCS [PWR1 or RPV [BWR]) inventory affecting core Loss of reactor vessel I RCS inventory affecting core decay heat removal decay heat removal capability. capabilities. to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability: 2) EAL #1 Quad Cities Station does not have a low, low, low RPV water level Cold Shutdown, Refueling 4,5 value stated within their Technical Specifications, all ESF functions come from the low, low value of -55.2inches, NEI 99-01 rev 6 developer notes for CSl Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): request the developer to use the low, low, low value for declaration of a SAE Note: The Emergency Director should declare the Site Area Emergency Note: when Containment Closure is not established, as such an RPV water level -65 promptly upon determining that 30 minutes time has been exceeded, or inches was selected as the threshold value since it is substantially above TAF will likely be exceeded.
| |
| The Emergency Director should declare the event promptly upon RPV water level of -142 inches and substantially below the low, low value of determining that the applicable time has been exceeded, or will likely be -55.2inches clearly indicating a worsening of the condition and the need to
| |
| : 1. a. CONTAINMENT CLOSURE not established. exceeded. escalate the emergency from an Alert condition.
| |
| AND
| |
| : b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-
| |
| : 1. With CONTAINMENT CLOSURE not established, RPV water level specific level). < -65 inches
| |
| : 2. a. CONTAINMENT CLOSURE established. OR AND 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 inches (TAF)
| |
| : b. (Reactor vessel/RCS [PWR] or RPV [BWRJ) level less than (site-specific level). OR
| |
| : 3. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be 3. a. RPV water level cannot be monitored for > 30 minutes monitored for 30 minutes or longer.
| |
| AND AND
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| " Table C3 indications of a sufficient magnitude to indicate core
| |
| " (Site-specific radiation monitor) reading greater than (site- uncovery.
| |
| specific value)
| |
| OR
| |
| " Erratic source range monitor indication [PWR]
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.
| |
| " UNPLANNED increase in (site-specific sump and/or tank levels) of sufficient magnitude to indicate core uncovery
| |
| " (Other site-specific indications) Table C3 Indications of RCS Leakage
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| * UNPLANNED Torus level rise*
| |
| * UNPLANNED vessel make up rate rise
| |
| * Observation of leakage or inventory loss
| |
| *Rise in level is attributed to a loss of RPV inventory.
| |
| Page 49 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CA1 CA6 Initiating Condition: ALERT Initiating Condition: D: No Change M Difference 1 Deviation Loss of (reactor vessel/RCS [PWRI or RPV [BWR]) inventory Loss of RPV inventory 1) Listed site specific levels, and sumps and tanks to ensure timely Operating Mode Applicability: Operating Mode Applicability: classification.
| |
| Cold Shutdown, Refueling 4,5 Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Alert promptly upon Note:
| |
| determining that 15 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be
| |
| : 1. Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as exceeded.
| |
| indicated by level less than (site-specific level).
| |
| : 2. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer 1. Loss of RPV inventory as indicated by level < - 59inches.
| |
| AND OR
| |
| : b. UNPLANNED increase in (site-specific sump and/or tank) levels due 2. a. RPV water level cannot be monitored for_> 15 minutes.
| |
| to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. AND
| |
| : b. Loss of RPV inventory per Table C3 indications.
| |
| Table C3 Indications of RCS Leakage
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| * UNPLANNED Torus level rise*
| |
| * UNPLANNED vessel make up rate rise
| |
| * Observation of leakage or inventory loss
| |
| *Rise in level is attributed to a loss of RPV inventory.
| |
| Page 50 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification Initiating Condition: UNUSUAL EVENT Cul Initiating Condition:
| |
| CU6 H No Change Difference Deviation UNPLANNED loss of (reactor vessel/RCS [PWRI or RPV [BWR]) inventory for UNPLANNED loss of reactor vessel / RCS inventory for 15 minutes or longer. 1) Described "a required lower limit" as a procedurally established lower limit, 15 minutes or longer. and Operating Mode Applicability:
| |
| Operating Mode Applicability: listed site specific sumps and tanks to ensure timely classification.
| |
| 4,5 Cold Shutdown, Refueling Emergency Action Level (EAL):
| |
| Example Emergency Action Levels: (1 or 2)
| |
| Note:
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be
| |
| * The Emergency Director should declare the event promptly upon exceeded. determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| : 1. UNPLANNED loss of reactor coolant results in (reactor vessel/RCS
| |
| [PWRj or RPV [BWRI) level less than a required lower limit for 15 minutes or longer. 1. UNPLANNED loss of reactor coolant results in the inability to restore and
| |
| : 2. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be maintain RPV water level above the procedurally established lower limit monitored. for > 15 minutes.
| |
| AND OR
| |
| : b. UNPLANNED increase in (site-specific sump and/or tank) levels. 2. a. RPV water level cannot be monitored AND
| |
| : b. Loss of RPV inventory per Table C3 indications.
| |
| Table C3 Indications of RCS Leakage
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| * UNPLANNED Torus level rise*
| |
| * UNPLANNED vessel make up rate rise
| |
| * Observation of leakage or inventory loss
| |
| *Rise in level is attributed to a loss of RPV inventory.
| |
| Page 51 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HG1 HG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition: D No Change M Difference F- Deviation HOSTILE ACTION resulting in loss of physical control of the facility. HOSTILE ACTION resulting in loss of physical control of the facility. 1) List site security shift supervision as Security Force.
| |
| : 2) Added descriptors to better explain each safety function and allow for a Operating Mode Applicability: Operating Mode Applicability: timely classification.
| |
| All 1,2,3,4,5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| : 1. a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift 1. A notification from the Security Force that a HOSTILE ACTION is occurring or has supervision). occurred within the PROTECTED AREA.
| |
| AND AND
| |
| : 2. a. ANY Table H1 safety function cannot be controlled or maintained.
| |
| : b. EITHER of the following: OR
| |
| : b. Damage to spent fuel has occurred or is IMMINENT
| |
| : 1. ANY of the following safety functions cannot be controlled or maintained.
| |
| Table H1 Safety Functions
| |
| * Reactivity control " Reactivity Control
| |
| * Core cooling [PWR] I RPV water level [BWR] (ability to shut down the reactor and keep it shutdown)
| |
| * RCS heat removal " RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| OR
| |
| : 2. Damage to spent fuel has occurred or is IMMINENT Page 52 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HSI HSI Initiating Condition: SITE AREA EMERGENCY Initiating Condition: No Change D Difference F Deviation HOSTILE ACTION within the Protected Area. HOSTILE ACTION within the Protected Area. 1) List site security shift supervision as Security Force.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| A HOSITLE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-security shift supervision). A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.
| |
| Page 53 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HAI HAI Initiating Condition: ALERT Initiating Condition: No Change FIDifference i1 Deviation HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack 1) List site security shift supervision as Security Force.
| |
| threat within 30 minutes. threat within 30 minutes.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):
| |
| : 1. A HOSTILE ACTION is occurring or has occurred within the OWNER
| |
| : 1. A validated notification from NRC of an aircraft attack threat < 30 minutes CONTROLLED AREA qs reported by the (site-specific security shift from the site.
| |
| supervision).
| |
| : 2. A validated notification from NRC of an aircraft attack threat within 30 OR minutes of the site.
| |
| : 2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.
| |
| Page 54 of 66
| |
| | |
| NEI 99-01 Rev 6 j Proposed EAL Justification HUI HUI Initiating Condition: UNUSUAL EVENT Confirmed SECURITY CONDITION or threat.
| |
| Initiating Condition: FD1 No Change Fq Difference F Deviation Confirmed SECURITY CONDITION or threat. 1) List site security shift supervision as Security Force.
| |
| Operating Mode Applicability: 2) Further described credible security threat through listing a site specific Operating Mode Applicability: procedure.
| |
| All 1,2, 3, 4, 5, D Example Emergency Action Levels: (1 or 2 or 3)
| |
| Emergency Action Level (EAL):
| |
| : 1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision). 1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.
| |
| : 2. Notification of a credible security threat directed at the site. OR
| |
| : 2. A validated notification from the NRC providing information of an aircraft
| |
| : 3. A validated notification from the NRC providing information of an aircraft threat.
| |
| threat.
| |
| OR
| |
| : 3. Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.
| |
| Page 55 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL [ Justification Initiating Condition: SITE AREA EMERGENCY HS6 Initiating Condition:
| |
| HS2 H* No Change Difference Deviation Inability to control a key safety function from outside the Control Room. Inability to control a key safety function from outside the Control Room. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would Operating Mode Applicability: Operating Mode Applicability: enter should such an event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate All 1, 2, 3,4, 5, D shutdown were to be listed,
| |
| : 2) Added descriptors to better explain each safety function and allow for a Example Emergency Action Levels: (1 and 2) Emergency Action Level (EAL): timely classification.
| |
| Note: The Emergency Director should declare the Site Area Emergency Note:
| |
| promptly upon determining that (site-specific number of minutes) has 3) Changed "An event" to" A Control Room evacuation" to remove confusion been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that if partial plant control was transferred to outside the control room with the the applicable time has been exceeded, or will likely be exceeded. control room still manned, due to testing or equipment failure.
| |
| : 1. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). 1 A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per
| |
| : 2. Control of ANY of the following key safety functions is not reestablished
| |
| * QCARP 0050-01, SB-1-llnjection with SSMP and Bringing the Unit to within (site-specific number of minutes). Cold Shutdown OR Reactivity control
| |
| * QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to
| |
| * Core cooling [PWR] / RPV water level [BWR] Cold Shutdown RCS heat removal OR
| |
| * ODA 0010-05, Plant Operation with the Control Room Inaccessible AND
| |
| : 2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.
| |
| Table H1 Safety Functions
| |
| * Reactivity Control (ability to shut down the reactor and keep it shutdown)
| |
| * RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| Page 56 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification HA6 HA2 Initiating Condition: ALERT Initiating Condition: D No Change F Difference FIDeviation Control Room evacuation resulting in transfer of plant control to alternate locations. Control Room evacuation resulting in transfer of plant control to alternate locations. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would enter should such an Operating Mode Applicability: Operating Mode Applicability: event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate shutdown were to be listed, All 1, 2, 3,4, 5, D
| |
| : 2) Changed "An event" to" A Control Room evacuation" to remove confusion if partial plant control was transferred to outside the control room with the control Example Emergency Action Levels: Emergency Action Level (EAL): room still manned, due to testing or equipment failure.
| |
| An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:
| |
| " QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| " QOA 0010-05, Plant Operation with the Control Room Inaccessible Page 57 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL j Justification HU4 HU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change EIDifference F Deviation FIRE potentially degrading the level of safety of the plant.
| |
| FIRE potentially degrading the level of safety of the plant. 1) Listed site specific list of plant rooms or areas that contain SAFETY SYSTEM equipment to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| : 2) Added additional note to easily direct the operator to escalation ICs All 1, 2, 3,4, 5, D Example Emergency Action Levels: (1 or 2 or 3 or 4)
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been Note:
| |
| exceeded, or will likely be exceeded.
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| : 1. a. A FIRE is NOT extinguished within 15-minutes of ANY of the
| |
| * Escalation of the emergency classification level would be via IC CA2 or MA5 following FIRE detection indications:
| |
| : 1. A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of the
| |
| * Report from the field (i.e., visual observation)
| |
| * Receipt of multiple (more than 1) fire alarms or indications following FIRE detection indications:
| |
| * Field verification of a single fire alarm
| |
| * Report from the field (i.e., visual observation)
| |
| AND
| |
| * Receipt of multiple (more than 1) fire alarms or indications
| |
| : b. The FIRE is located within ANY of the following plant rooms or
| |
| * Field verification of a single fire alarm areas: Table H2 Vital Areas (site-specific list of plant rooms or areas)
| |
| * Reactor Building (when inerted the Drywell is exempt)
| |
| * Main Control Room Envelope
| |
| : 2. a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).
| |
| * Unit and Shared Emergency Diesel Generator Rooms AND
| |
| * 4KV Switchgear Area
| |
| : b. The FIRE is located within ANY of the following plant rooms or
| |
| * Battery Rooms areas:
| |
| * RHR Service Water Vaults (site-specific list of plant rooms or areas)
| |
| * Turbine Building Cable Tunnel AND
| |
| * Cribhouse
| |
| : c. The existence of a FIRE is riot verified within 30-minutes of alarm OR receipt.
| |
| : 2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of a FIRE).
| |
| : 3. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the AND plant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. b. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.
| |
| OR
| |
| : 4. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the 3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-plant Protected Area] PROTECTED AREA that requires firefighting minutes of the initial report, alarm or indication.
| |
| support by an offsite fire response agency to extinguish. OR
| |
| : 4. A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
| |
| Page 58 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HU2 Initiating Condition: UNUSUAL EVENT Initiating Condition:
| |
| HU4 D No Change
| |
| * Difference F Deviation Seismic event greater than OBE levels. 1) Used Alternate developer notes allowed wording since specific Control Room Seismic event greater than OBE levels.
| |
| indication of a seismic event > OBE is not available.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability: 2) Added additional compensatory thresholds based on the guidance provided in US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear All Power Plant Operator Earthquake Actions Appendix A, also added a 15 min time 1,2, 3,4, 5, D limit to ensure timely classification when these additional thresholds may not be Example Emergency Action Levels: Emergency Action Level (EAL): readily available.
| |
| Note: Escalation of the emergency classification level would be via IC CA2 or Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: MA5 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
| |
| : a. (site-specific indication that a seismic event met or exceeded OBE limits) For emergency classification if EAL 2.b is not able to be confirmed, then will determine classification within 15 mins of the event the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in 1 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.
| |
| Page 59 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification
| |
| -F -I H HA5 A5 Initiating Condition: ALERT Initiating Condition: D No Change ] Difference D eviation Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or 1) Listed plant specific rooms and areas with entry related mode Gaseous release impeding access to equipment necessary for normal shutdown. applicability to ensure timely classification.
| |
| plant operations, cooldown or shutdown.
| |
| : 2) The Operating Mode Applicability of this EAL has been revised from All Operating Mode Applicability: Modes to modes 3 and 4 due to the mode applicability of the areas of Operating Mode Applicability: concern in Table H-3.
| |
| 3,4 All Emergency Action Level (EAL):
| |
| Example Emergency Action Levels:
| |
| Note:
| |
| Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no If the equipment in the listed room or area was already inoperable, or out of service, before emergency classification is warranted. the event occurred, then no emergency classification is warranted.
| |
| : 1. a. Release of a toxic, corrosive, asphyxiant or flammable gas
| |
| : 1. Release of a toxic, corrosive, asphyxiant or flammable gas in a Table H3 area.
| |
| into any of the following plant rooms or areas:
| |
| (site-specific list of plant rooms or areas with entry-related Table H3 mode applicability identified) Areas with Entry Related Mode Applicability Entry Related AND Area Unit Mode Applicability
| |
| : b. Entry into the room or area is prohibited or impeded. Reactor Building
| |
| " First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| " First Floor South Wall 2
| |
| " Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 AND
| |
| : 2. Entry into the room or area is prohibited or impeded.
| |
| Page 60 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HU3 HU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change FIDifference F- Deviation Hazardous Event Hazardous Event 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 Operating Mode Applicability: EAL UE values.
| |
| Operating Mode Applicability:
| |
| All 2) Changed the word "needed" to "required by Technical Specifications" in the EAL 1,2, 3,4, 5, D to be consistent with terminology used by operators and minimize confusion.
| |
| Example Emergency Action Levels: (1 or 2 or 3 or 4)
| |
| Emergency Action Level (EAL): 3) Added additional note to easily direct the operator to escalation ICs Note: EAL #3 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
| |
| Note:
| |
| I. A tornado strike within the PROTECTED AREA.
| |
| * EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
| |
| : 2. Internal room or area flooding of a magnitude sufficient to require manual
| |
| * Escalation of the emergency classification level would be via IC CA2 or or automatic electrical isolation of a SAFETY SYSTEM component needed MA5 for the current operating mode.
| |
| : 1. Tornado strike within the PROTECTED AREA.
| |
| : 3. Movement of personnel within the PROTECTED AREA is impeded due to OR an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release). 2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by
| |
| : 4. A hazardous event that results in on-site conditions sufficient to prohibit Technical Specifications for the current operating mode.
| |
| the plant staff from accessing the site via personal vehicles. OR
| |
| : 5. (Site-specific list of natural or technological hazard events) 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 River level, as indicated by EITHER:
| |
| : a. High river level > 594 ft.
| |
| OR
| |
| : b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam #14 has failed.
| |
| Page 61 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification HG7 HG7 Initiating Condition: GENERAL EMERGENCY Initiating Condition: W No Change EIDifference F Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. declaration of a General Emergency.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT Other conditions exist which in the judgment of the Emergency Director indicate substantial core degradation or melting with potential for loss of containment that events are in progress or have occurred which involve actual or IMMINENT integrity or HOSTILE ACTION that results in an actual loss of physical control of substantial core degradation or melting with potential for loss of containment the facility. Releases can be reasonably expected to exceed EPA Protective Action integrity or HOSTILE ACTION that results in an actual loss of physical control of Guideline exposure levels off-site for more than the immediate site area. the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area..
| |
| Page 62 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: SITE AREA EMERGENCY Initiating Condition: LSi No Change [JDifference Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. declaration of a Site Area Emergency.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Other conditions exist which in the judgment of the Emergency Director indicate Other conditions exist which in the judgment of the Emergency Director indicate that major that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION fainu that results in intentional damage or malicious acts; (1) toward site personnel or failures of plant functions needed for protection of the public or HOSTILE ACTION equipment that could lead to the likely failure of or; (2) that prevent effective that results in intentional damage or malicious acts; (1) toward site personnel or access to equipment needed for the protection of the public. Any releases are not equipment that could lead to the likely failure of or; (2) that prevent effective expected to result in exposure levels which exceed EPA Protective Action access to equipment needed for the protection of the public. Any releases are not Guielinetedtoresultin exposure levels be thextebuda, Eexpected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. Guideline exposure levels beyond the site boundary.
| |
| Page 63 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: ALERT Initiating Condition: Lhi No Change [J Difference Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. declaration of an Alert.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Other conditions exist which in the judgment of the Emergency Director indicate Other conditions exist which in the judgment of the Emergency Director indicate that events degradation substantial are in progress or have of the level occurred of safety which involve of the plant or an actual orevent a security potential that that events are in progress or have occurred which involve an actual or potential substanvolvdesgpradatn lfe threatenln rkof safetyohe peonn ordamagetosityev 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 siteinovspballfetrtnngiktoiepronlordmetoie equipment because of HOSTILE ACTION. Any releases are expected to be involves probable life threatening risk to site personnel or damage to site limited to small fractions of the EPA Protective Action Guideline exposure levels, equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
| |
| Page 64 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HU7 HU7 r m m Initiating Condition: UNUSUAL EVENT Initiating Condition: L No Change Difference Deviation Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT. Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| All 1,2, 3,4, 5, D Example Emergency Action Levels:
| |
| Other conditions exist which in the judgment of the Emergency Director indicate Emergency Action Level (EAL):
| |
| that events are in progress or have occurred which indicate a potential Other conditions exist which in the judgment of the Emergency Director indicate degradation of the level of safety of the plant or indicate a security threat to facility that events are in progress or have occurred which indicate a potential protection has been initiated. No releases of radioactive material requiring offsite degradation of the level of safety of the plant or indicate a security threat to facility response or monitoring are expected unless further degradation of safety systems protection has been initiated. No releases of radioactive material requiring offsite occurs. response or monitoring are expected unless further degradation of safety systems occurs.
| |
| Page 65 of 66
| |
| | |
| NEI 99-01 Rev 6 [ Proposed EAL [ Justification E-HU1 Initiating Condition: UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY.
| |
| Initiating Condition:
| |
| E-HUI W-] No Change 1 Difference i Deviation Damage to a loaded cask CONFINEMENT BOUNDARY. 1) Listed 2x the site specific cask specific allowable radiation level as per Certificate of Compliance No. 1014 Appendix A Section 5.7.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-technical specification allowable radiation level) on the surface of the spent fuel contact radiation reading:
| |
| cask.
| |
| * > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR
| |
| * > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts.
| |
| Page 66 of 66
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 Initiating Condition:
| |
| Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRem TEDE or 5000 mRem thyroid CDE.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| * Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond 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.
| |
| Month 20XX QC 3-33 EP-AA-1006 (Revision XX)
| |
| | |
| (1"2ri ifina Annov I=:vgm-In N"Aaln-r flu .~.a~
| |
| ~- ri Ci*ic~c
| |
| -
| |
| * t*~o Anncv 5-s. .. . n ~ . ~. . . . - ~. - ~A.
| |
| RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 (cont)
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG1
| |
| : 2. EP-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 4. QGA 400 Radioactivity Release Control Month 20XX QC 3-34 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RS1 Initiating Condition:
| |
| Release of gaseous radioactivity resulting in offsite dose greater than 100 mRem TEDE or 500 mRem thyroid CDE.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| 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.
| |
| Month 20XX QC 3-35 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RSI (cont)
| |
| 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-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-36 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| * Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
| |
| : a. > 10 mRem TEDE OR
| |
| : b. > 50 mRem CDE Thyroid OR Month 20XX QC 3-37 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
| |
| : a. 10 mRem TEDE for 60 minutes of exposure OR
| |
| : b. 50 mRem CDE Thyroid for 60 minutes of exposure OR
| |
| : 4. Field survey results at or beyond the site boundary indicate EITHER:
| |
| : a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for > 60 minutes.
| |
| OR
| |
| : b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Basis:
| |
| This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of 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.
| |
| Escalation of the emergency classification level would be via IC RS1.
| |
| Month 20XX QC 3-38 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA1
| |
| : 2. CY-QC-120-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. QOP 2000-24, Discharging to the River from the River Discharge Tank using the Waste Surge Pump
| |
| : 6. QOP 2000-25, Discharging to the River from the River Discharge Tank using the River Discharge Pump
| |
| : 7. CY-QC-1 20-729, Liquid Effluent Alarm Setpoints
| |
| : 8. QCOA 1700-02, High Radiation detected on Eberline Radiation Monitoring System
| |
| : 9. QCAN 912-5 C-6, Radwaste High Rad.
| |
| : 10. QCAN 901(2)-3-G-1, Liquid Process Rad. Monitor High Radiation
| |
| : 11. CY-QC-1 20-735, Main Chimney & Reactor Vent Noble Gas Release Rate Action Levels
| |
| : 12. QCOA 1700-01, Abnormal Chimney Radiation
| |
| : 13. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 14. QGA 400 Radioactivity Release Control Month 20XX QC 3-39 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 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, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
| |
| : 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for > 60 minutes.
| |
| " Radwaste Effluent Monitor 1/2-1799-01 OR
| |
| " Discharge Permit specified monitor OR
| |
| : 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
| |
| Basis:
| |
| This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
| |
| Month 20XX QC 3-40 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RUl (cont)
| |
| Basis (cont):
| |
| 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.
| |
| EAL #1 Basis This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
| |
| The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.
| |
| EAL #2 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous effluent pathways.
| |
| EAL #3 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 RA1.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU1
| |
| : 2. CY-QC-120-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-41 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, 6, 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.
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG2 Month 20XX QC 3-42 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex OuadCites AnexExelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS2 Initiating Condition:
| |
| Spent fuel pool level at (site-specific Level 3 description).
| |
| Operating Mode Applicability:
| |
| 1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, AS2 Month 20XX QC 3-43 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 Initiating Condition:
| |
| Significant lowering of water level above, or damage to, irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1, 2, 3,4, 5, D Emergency Action Level (EAL):
| |
| : 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.
| |
| OR
| |
| : 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R1 Radiation Monitor reading >1000 mRem/hr.
| |
| OR
| |
| : 3. Lowering of spent fuel pool level to (site specific Level 2 value).
| |
| Table R1 Fuel Handling Incident Radiation Monitors 0 1(2) 1705-16A Fuel Pool Rad Monitor
| |
| * 1(2) 1705-16B Fuel Pool Rad Monitor Basis:
| |
| REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.
| |
| IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
| |
| CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-44 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 (cont)
| |
| Basis (cont):
| |
| 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 an 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 Ror C ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA2
| |
| : 2. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity
| |
| : 3. QCAN 901(2)-3 B-1 Refuel Floor Hi Radiation
| |
| : 4. QCAN 901(2)-3 G-16/H-16 Fuel Pool Channel A/B Hi Radiation
| |
| : 5. QCIS 1700-07 Reactor Building Ventilation and Fuel Pool Radiation Monitoring Calibration and Functional Test Month 20XX QC 3-45 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 Initiating Condition:
| |
| UNPLANNED loss of water level above irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| : 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
| |
| " Refueling Cavity water level < 282 in. (Upper Wide range simulated signal).
| |
| OR
| |
| " Spent Fuel Pool water level < 19 ft. above the fuel (<- 4 ft. indicated level).
| |
| OR
| |
| " Indication or report of a drop in water level in the REFUELING PATHWAY.
| |
| AND
| |
| : b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R1.
| |
| Table R1 Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Mon
| |
| * 1(2) 1705-16B Fuel Pool Rad Mon Month 20XX QC 3-46 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 (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.
| |
| 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 an 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU2
| |
| : 2. QCOP 0201-13 Reactor Level Upper Wide Range Reference Leg Extension Use and Control
| |
| : 3. Technical Specifications 3.7.8 Spent Fuel Storage Pool Water Level
| |
| : 4. Technical Specifications 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel
| |
| : 5. QCAN 901(2)-4 B-24 FUEL POOL STORAGE HI/LO LEVEL
| |
| : 6. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity Month 20XX QC 3-47 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Exelon Nnncl*ar Nuclea~r Duasd Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RA3 Initiating Condition:
| |
| Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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 R2 areas:
| |
| Table R2 Areas Requiring Continuous Occupancy
| |
| " Main Control Room (Unit 1 ARM Station #22)
| |
| * Central Alarm Station - (by survey)
| |
| OR
| |
| : 2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:
| |
| Table R3 Arcas with ENtr Related -, Mode Appliabiity A~eEntry Rclatcd Mode
| |
| __Appllr~abillity.
| |
| Reactor Building* Mode 3 and-4 Turbine Building
| |
| * Mode 3
| |
| *Areas required to establish shutdown GEcoolin Month 20XX QC3-48 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Table R3 Areas with Entry Related Mode Applicability Area Unit Entry Related Mode Applicability Reactor Building
| |
| " First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 Month 20XX QC 3-49 EP-AA-1 006 (Revision XX)
| |
| | |
| FYeInn NHP.I*_*r Ouadr Cities~ AnnAex Fvyinnn Nmriar~I RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.
| |
| This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
| |
| Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.
| |
| For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in 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.
| |
| " 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.).
| |
| Month 20XX QC 3-50 EP-AA-1006 (Revision XX)
| |
| | |
| l:YpInn N.r.l*n r Dua~d Cities Annex FyDinn M..r~iozr RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RA3 (cont)
| |
| Basis (cont):
| |
| " 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. QCOP 1800-1 Operation of ARM Indicator/Trip Units
| |
| : 3. UFSAR Section 3.2
| |
| : 4. General Arrangement Drawings M-5, 6, 8 and 10 Month 20XX QC 3-51 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU3 Initiating Condition:
| |
| Reactor coolant activity greater than Technical Specification allowable limits.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| : 1. Offgas system radiation monitor HI-HI alarm.
| |
| OR
| |
| : 2. Specific coolant activity > 4.0 uCi/gm Dose equivalent 1-131.
| |
| Basis:
| |
| This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.
| |
| Conditions that cause the specified monitor to alarm that are not related to fuel clad degradation should not result in the declaration of an Unusual Event.
| |
| This EAL addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity.
| |
| An Unusual EVent is only warranted when actual fuel clad damage is the cause of the elevated coolant sample activity (as deteFrmined by laboratory confiFrmation). Fuel cla damage should be assumed to be the cause o)f elevated Reactor Coolant activity unless anoher c ause is known.
| |
| Escalation of the emergency classification level would be via I(s FA1 or the Recognition Category R ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU3
| |
| : 2. Technical Specifications 3.4.6
| |
| : 3. Technical Specifications 3.7.6
| |
| : 4. QCAN 901(2)-3 C-2 OFF GAS HIGH-HIGH RADIATION Month 20XX QC 3-52 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:
| |
| Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the General Emergency classification level each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-53 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:
| |
| Loss or Potential Loss of ANY two barriers.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the Site Area Emergency classification level, each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-54 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:
| |
| ANY Loss or ANY Potential Loss of EITHER Fuel Clad or RCS.
| |
| Operating Mode Applicability:
| |
| 1,2,3 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 Month 20XX QC 3-55 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn N.r.lpar r~uad Cities Annex FYAlnn Nucler~I RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:
| |
| RCS Activity Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Coolant activity > 300 uCi/gm Dose Equivalent 1-131.
| |
| Basis:
| |
| This threshold indicates that RCS radioactivity concentration is greater than 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.
| |
| 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
| |
| : 2. NF-AA-430, Failed Fuel Action Plan Month 20XX QC 3-56 EP-AA-1006 (Revision XX)
| |
| | |
| Duadr Cities~ Annex Exelon Nuceiair RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:
| |
| RPV Water Level 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 > -142 inches (TAF)
| |
| OR
| |
| : 3. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.
| |
| This is identified in the BWROG EOPs/SAMGs 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 Month 20XX QC 3-57 EP-AA-1006 (Revision XX)
| |
| | |
| fl"!arl rifiac Annov IPvnlnn hn~rlaar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)
| |
| Basis (cont):
| |
| RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
| |
| The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-58 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Drywell radiation monitor reading > 6.65 E+02 R/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 300 ý0Ci/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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. Core Damage Assessment Methodology (CDAM)
| |
| Month 20XX QC 3-59 EP-AA-1006 (Revision XX)
| |
| | |
| (1".qd rifloa Annoy I:::YlrIn Miaerla-r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-60 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. RPV water level cannot be restored and maintained > -142 inches (TAF)
| |
| OR
| |
| : 2. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.
| |
| The RPV water level threshold is the same as Fuel Clad 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 Month 20XX QC 3-61 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex OuadCites AnexExelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)
| |
| Basis (cont):
| |
| fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 500-4 RPV Flooding Month 20XX QC 3-62 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:
| |
| Primary Containment Pressure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell pressure >2.5 psig.
| |
| AND
| |
| : 2. Drywell pressure rise is due to RCS leakage.
| |
| Basis:
| |
| The > 2.5 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.
| |
| The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.
| |
| The release of mass from the RCS due to the as-designed/expected operation of any relief valve does not warrant an emergency classification.
| |
| A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| There is no Potential Loss threshold associated with Primary Containment Pressure.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 200 Primary Containment Control Month 20XX QC 3-63 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:
| |
| RCS Leak Rate Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC 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 > QGA 300 Maximum Normal operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300 Maximum Normal 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 on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.
| |
| 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.
| |
| Month 20XX QC 3-64 EP-AA-1006 (Revision XX)
| |
| | |
| n"nrl Mae Annav F~vale~n NMgirInr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis (cont):
| |
| 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 safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the Torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
| |
| Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
| |
| A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
| |
| The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.
| |
| In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.
| |
| An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold
| |
| #1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
| |
| Month 20XX QC 3-65 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. M-13 Main steam piping
| |
| : 3. UFSAR 5.2.5
| |
| : 4. QCOA 0201-01 Rev 016, Increasing Drywell Pressure
| |
| : 5. QOA 900-4 A-17 900-4 A-17 Annunciator
| |
| : 6. QCOS 1600-07 Reactor Coolant Leakage In The Drywell
| |
| : 7. QGA 300 Secondary Containment Control Month 20XX QC 3-66 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:
| |
| Primary Containment radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell radiation monitor reading> I OOR/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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. CaIc. EP-EAL-0611 Month 20XX QC 3-67 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:
| |
| 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.
| |
| 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 Month 20XX QC 3-68 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-69 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn FvAlnn Nuclear NuclerIzr Duad Cities Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:
| |
| Primary Containment 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 > 56 psig and rising.
| |
| OR
| |
| : 4. a. Drywell or torus hydrogen concentration > 6%.
| |
| AND
| |
| : b. Drywell or torus oxygen concentration > 5%.
| |
| OR
| |
| : 5. Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| Loss Threshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
| |
| These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.
| |
| Month 20XX QC 3-70 EP-AA-1006 (Revision XX)
| |
| | |
| n"nd ritia4z Annoy IPvle, n Ih, r~lanr Am i~.rI Citi~~ Ann~v FvaIr~n MmirI~2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)
| |
| Basis (cont):
| |
| Potential Loss Threshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.
| |
| Potential Loss 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. UFSAR Fig. 6.2-16a
| |
| : 3. UFSAR Section 15.6
| |
| : 4. UFSAR 6.2.1.1
| |
| : 5. QGA 200, Primary Containment Control
| |
| : 6. Quad Cities PSTG Section 5, Primary Containment Control Month 20XX QC 3-71 EP-AA-1 006 (Revision XX)
| |
| | |
| ExAlnn Nuclear Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS
| |
| : 1. Drywell radiation monitor reading > 1.55 E+03 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-1 228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20%
| |
| in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. Core Damage Assessment Methodology (CDAM)
| |
| Month 20XX QC 3-72 EP-AA-1006 (Revision XX)
| |
| | |
| Oiiad Cities Annex Exelon Nuclear I=xelon Nucle~ar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:
| |
| Primary Containment Isolation Failure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 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 > QGA 300, Maximum Safe operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300, Maximum Safe operating levels.
| |
| 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, HPCI or RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.
| |
| Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.
| |
| Month 20XX QC 3-73 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Month 20XX QC 3-74 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)
| |
| Basis (cont):
| |
| Main Turbine Control Valve closure) may indicate that a, primary system is discharging into the Reactor Building.
| |
| In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.
| |
| 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. QGA 200 Primary Containment Control
| |
| : 3. QGA 200-5 Hydrogen Control
| |
| : 4. QCOP 1600-13 Post-Accident Venting of the Primary Containment
| |
| : 5. QGA 300 Secondary Containment Control Month 20XX QC 3-75 EP-AA-1006 (Revision XX)
| |
| | |
| A 112 rl Cificn Annov I::v-Inln N, n-I-nr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-76 EP-AA-1006 (Revision XX)
| |
| | |
| fl"!mfl ('ifiac Annov Pvalewn N~irlanr 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. EITHER of the following:
| |
| : a. Restoration of at least one unit ECCS bus in < 1 hour is not likely.
| |
| OR
| |
| : b. RPV water level cannot be restored and maintained > -166 inches.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.
| |
| This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat 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.
| |
| Month 20XX QC 3-77 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellan Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)
| |
| Basis (cont):
| |
| Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
| |
| The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
| |
| The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. QGA 100 RPV Control Month 20XX QC 3-78 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Nuclear QuadCites AnexExelon 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.
| |
| Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
| |
| Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.
| |
| Month 20XX QC 3-79 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon NuclIear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-80 EP-AA-1006 (Revision XX)
| |
| | |
| rliinfl riflac Annav Pvi:=nn klielanr A. .~ w4Citig~ Annoy Fvoinn Mm me102r RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit Auxiliary Transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers 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 MUl.
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| * A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
| |
| Month 20XX QC 3-81 EP-AA-1006 (Revision XX)
| |
| | |
| Exallon Eeo Nuclear ula Quad Cities Annex RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)
| |
| Basis (cont):
| |
| * A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-82 EP-AA-1006 (Revision XX)
| |
| | |
| 01iad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3 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 unit ECCS busses 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. (e.g. unit cross-tie breakers)
| |
| Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
| |
| Escalation of the emergency classification level would be via IC MAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout Month 20XX QC 3-83 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.
| |
| AND
| |
| : 3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.
| |
| AND
| |
| : 4. ALL AC and Vital DC power sources have been lost for >15 minutes.
| |
| 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 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 EALs are met.
| |
| Month 20XX QC 3-84 EP-AA-1006 (Revision XX)
| |
| | |
| l=x*Inn N mmr.lpnr r3uad Cities Annex Fxelon Nue-lanir RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG8
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. Technical Specifications 3.8.4 and B3.8.4
| |
| : 10. UFSAR Section 8.3.2
| |
| : 11. QOP 6900-02 125 VDC Electrical System
| |
| : 12. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-85 EP-AA-1006 (Revision XX)
| |
| | |
| QUad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2 Initiating Condition:
| |
| Loss of all vital DC power for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| Voltage is < 105 VDC on 125 VDC battery busses #1 and #2 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS8
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-86 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 > 5%.
| |
| AND
| |
| : 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
| |
| AND
| |
| : 3. EITHER of the following conditions exist:
| |
| * RPV water level cannot be restored and maintained > -166 inches.
| |
| OR
| |
| * Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| 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.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Escalation of the emergency classification level would be via IC RG1 or FGI.
| |
| Month 20XX QC 3-87 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex I=x*lon Nuclear OuadCites AnexExelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 200 Primary Containment Control Month 20XX QC 3-88 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear 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:
| |
| 1,2 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 > 5%.
| |
| AND
| |
| : 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.
| |
| Basis:
| |
| This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.
| |
| Month 20XX QC 3-89 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| Month 20XX QC 3-90 EP-AA-1006 (Revision XX)
| |
| | |
| O.,ad Cities Annex Exelnn Nuc:l*r I=x*lon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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
| |
| > 5%.
| |
| 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
| |
| > 5%.
| |
| 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 control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
| |
| Month 20XX QC 3-91 EP-AA-1006 (Revision XX)
| |
| | |
| IFvAlnn MhirnIl~ar RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.
| |
| The plant response to the failure of an automatic or manual reactor 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 control 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.
| |
| Month 20XX QC 3-92 EP-AA-1006 (Revision XX)
| |
| | |
| I*xAInn Nuclear Oijad Cities Annex Exelon NuclerIsr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. Technical Specifications Table 3.3.1.1-1
| |
| : 5. Technical Specification 3.3.1.3
| |
| : 6. Technical Specification Bases 3.3.1.1 and 3.3.1.3 Month 20XX QC 3-93 EP-AA-1 006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| " Reactor Power
| |
| " RPV Water Level
| |
| " RPV Pressure
| |
| " Drywell Pressure
| |
| " Torus Level
| |
| " Torus Temperature AND
| |
| : 2. ANY Table M2 transient in progress.
| |
| Table M2 Significant Transients
| |
| " Turbine Trip
| |
| " Reactor Scram
| |
| " ECCS Activation
| |
| * Recirc. Runback > 25% Reactor Power Change
| |
| * Thermal Power oscillations > 10% Reactor Power Change Month 20XX QC 3-94 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear 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 Month 20XX QC 3-95 EP-AA-1006 (Revision XX)
| |
| | |
| rl"nrl ifina Annoy IPyalen N~rlalr Am
| |
| -- i~id
| |
| - - -Ci*iaQ
| |
| *U - Ann~v F 5~*
| |
| * S ~fl Fv~InnS M.urIa~r U- ~
| |
| RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| * Reactor Power
| |
| * RPV Water Level
| |
| * RPV Pressure
| |
| * Drywell Pressure
| |
| * Torus Level
| |
| * Torus Temperature Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
| |
| As used in this EAL, an "inability to monitor" means that values for 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.
| |
| Month 20XX QC 3-96 EP-AA-1006 (Revision XX)
| |
| | |
| Mad Cities Annex Nuclear OuadCites AnexExelon 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 Month 20XX QC 3-97 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 Initiating Condition:
| |
| 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.
| |
| 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 Month 20XX QC 3-98 EP-AA-1006 (Revision XX)
| |
| | |
| n"nrl rifina Annoy l::vg* In Ni,,el-anr A u12r4
| |
| -- - - -Citi~~c
| |
| *~ ~Annav
| |
| *~ * ~fl Fv~Inn
| |
| ~ . . - *vIa~r
| |
| * ~. . Nh ~
| |
| RECOGNITION CATEGORY SYSTEM MALFUNCTIONS events may require a post-event inspection to determine if the attributes of an explosion are present.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. These are typically systems classified as safety-related.
| |
| Month 20XX QC 3-99 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exallon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| This EAL #2a 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.
| |
| This 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 is not 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 SA9 Month 20XX QC 3-100 EP-AA-1006 (Revision XX)
| |
| | |
| Dijad Cities Annex OuadCitesRECOGNITION AnexExelon CATEGORY Nuclear SYSTEM MALFUNCTIONS 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.
| |
| 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 Month 20XX QC 3-101 EP-AA-1006 (Revision XX)
| |
| | |
| Omiad Cities Annex Exelon Nuclear Exelon Nuclea~r RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)
| |
| Basis (cont):
| |
| 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 Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| 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. QCOS 1600-07 Reactor Coolant Leakage in the Drywell
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QCOA 0201-01 Increasing Drywell Pressure Month 20XX QC 3-102 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Initiating Condition:
| |
| Loss of all On-site or Off-site communications capabilities.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency 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 Radio X Plant Page X Sound Powered Phones X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite 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.).
| |
| Month 20XX QC 3-103 EP-AA-1 006 (Revision XX)
| |
| | |
| I::xelnn Nuclear O.,ad Cities Annex Exellnn NuclerIzr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis This EAL addresses a total loss of the communications methods used in support of routine plant operations.
| |
| EAL #2 Basis This EAL 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) Form.
| |
| EAL #3 Basis This EAL 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-104 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. Loss of ALL offsite AC power to unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| 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.
| |
| Month 20XX QC 3-105 EP-AA-1006 (Revision XX)
| |
| | |
| I=xplnn NHeI*_*r Quad Cities Annex Exellnn NuclerIsr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA2
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "ACTURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-106 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS cul Initiating Condition:
| |
| Loss of all but one AC power source to emergency buses for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. AC power capability to unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit auxiliary transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers 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.
| |
| Month 20XX QC 3-107 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CUl (cont)
| |
| Basis (cont):
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| * 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 CAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6 CU2
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-108 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:
| |
| Hazardous event affecting SAFETY SYSTEM required for the current operating mode.
| |
| Operating Mode Applicability:
| |
| 4,5 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.
| |
| 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 Month 20XX QC 3-109 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex I=xplnn N HP.Ip_:* r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS events may require a post-event inspection to determine if the attributes of an explosion are present.
| |
| Month 20XX QC 3-110 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA6 Month 20XX QC 3-111 EP-AA-1 006 (Revision XX)
| |
| | |
| FY*_lnn Nnnr.l*r flm,21 CitiA;s Annex ExpInn NMipar~I RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:
| |
| Loss of Vital DC power 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.
| |
| Voltage is < 105 VDC on required 125 VDC battery busses #1 and #2 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 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.
| |
| Month 20XX QC 3-112 EP-AA-1006 (Revision XX)
| |
| | |
| Omar] ritioc Annoy FI:vlrln M,,ril r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CU4
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-113 EP-AA-1006 (Revision XX)
| |
| | |
| (3"2rl Mae Annoy I:volenn nMmr-l-r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:
| |
| Loss of all onsite or offsite communications capabilities.
| |
| Operating Mode Applicability:
| |
| 4,5, 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 C1 Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X Sound Pewcred Phones X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite communications capabilities.
| |
| While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Outside 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.
| |
| Month 20XX QC 3-114 EP-AA-1 006 (Revision XX)
| |
| | |
| Mind ritioc Annoy Nig-lanr P-mairnnM 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-115 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 Initiating Condition:
| |
| Inability to maintain the plant in cold shutdown.
| |
| 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-.
| |
| " 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 > 212OF duc to loss of decay heat removal for > Table C2 duration.
| |
| Table C2 RCS Heat-up Duration Thresholds RCS Containment Closure Heat-up Status Status Duration Intact Not Applicable 60 minutes*
| |
| Not Intact Established 20 minutes*
| |
| 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-te loss of decay heat removal.
| |
| 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.
| |
| Month 20XX QC 3-116 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS 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.).
| |
| Month 20XX QC 3-117 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Month 20XX QC 3-118 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. OU-AA-103 Shutdown Safety
| |
| : 6. QCOA 1000-02 Loss of Shutdown Cooling
| |
| : 7. QGA 100 RPV Control
| |
| : 8. QGA 100 RPV Control Detail A
| |
| : 9. QCGP 1-1 Normal Unit Startup
| |
| : 10. QCIS 0600-01 Unit One Division 1 Reactor Pressure 0 to 1200 psig Indication Calibration Month 20XX QC 3-119 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Ouad Cties nnex xelon Nuc~lear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 Initiating Condition:
| |
| UNPLANNED rise in RCS temperature 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.
| |
| " 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 removal.
| |
| 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.).
| |
| A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
| |
| Month 20XX QC 3-120 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn N Hc.l*_*r Duad Cities~ Annex Fxellnn Nui~clear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 (cont)
| |
| Basis (cont):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, CU3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. QGA 100, RPV Control
| |
| : 4. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 5. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 6. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge
| |
| : 7. QCOA 1000-02 Loss of Shutdown Cooling Month 20XX QC 3-121 EP-AA-1 006 (Revision XX)
| |
| | |
| Exelon Nuclear Eeo ula Duadr Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Initiating Condition:
| |
| Loss of RPV inventory affecting fuel clad integrity with containment challenged.
| |
| 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 a. RPV water level < -142 inches (TAF) for > 30 minutes.
| |
| AND
| |
| : b. Any Containment Challenge Indication (Table C4)
| |
| OR
| |
| : 2. a. RPV water level unknewn 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
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-122 EP-AA-1006 (Revision XX)
| |
| | |
| I=xelnn Exelnn Nucle_*r Nucler~I Omiad Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| Table C4 Containment Challenge Indications
| |
| " Primary Containment Hydrogen Concentration > 6% and Oxygen > 5%
| |
| " UNPLANNED rise in containment pressure
| |
| , CONTAINMENT CLOSURE not established*
| |
| " ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.
| |
| * if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.
| |
| 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.
| |
| Month 20XX QC 3-123 EP-AA-1006 (Revision XX)
| |
| | |
| I=xAInn N, m.l*_*r Dua~d Cities Annex FvAinn Nur~ipar RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis (cont):
| |
| 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-124 EP-AA-1006 (Revision XX)
| |
| | |
| tl"nei rifine Annoy I-on-Itn M,,l"A-r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CG1
| |
| : 2. QGA 100, RPV Control
| |
| : 3. Technical Specifications 3.3.1
| |
| : 4. Technical Specifications 3.6.1.1
| |
| : 5. Technical Specifications 3.6.4.1
| |
| : 6. QGA-200-5, Hydrogen Control
| |
| : 7. UFSAR 6.2.1.1
| |
| : 8. QGA 300 Secondary Containment Control
| |
| : 9. EP-EAL-0501, Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Month 20XX QC 3-125 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 Initiating Condition:
| |
| Loss of RPV inventory affecting core decay heat removal capability.
| |
| 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. With CONTAINMENT CLOSURE not established, RPV water level
| |
| < -65 inches OR
| |
| : 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 inches (TAF).
| |
| OR
| |
| : 3. a. RPV water level cannot be monitoredu'nknown 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
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-126 EP-AA-1006 (Revision XX)
| |
| | |
| I=xe.lnn N Hc.lp_*r Quad Cities Annex Exeiln NIuiclar RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 (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.
| |
| 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.b and 2.b 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 tank levels. Sump and/or tank 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 RG1.
| |
| Month 20XX QC 3-127 EP-AA-1006 (Revision XX)
| |
| | |
| Oijad Cities Annex ExeIon Nuclear I=x*lon Nucler!I RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CS1
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. QGA 100, RPV Control
| |
| : 6. Technical Specifications 3.3.1
| |
| : 7. Technical Specifications Table 3.3.3.1-1
| |
| : 8. Technical Specifications 3.3.5.1
| |
| : 9. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 10. Technical Specifications 3.4.4
| |
| : 11. UFSAR 5.2.5
| |
| : 12. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 13. QOA 900-4 A-17, Annuciator Response Month 20XX QC 3-128 EP-AA-1006 (Revision XX)
| |
| | |
| IFxAlnn NucleIasr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:
| |
| Loss of RPV 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 < - 59 inches.
| |
| OR
| |
| : 2. a. RPV water level cannot be monitored unkwn-;,for > 15 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 in 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 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.
| |
| Month 20XX QC 3-129 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis A lowering of water level below -59 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 decrease in water level will lead to core uncovery.
| |
| 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). An 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 tank levels. Sump and/or tank 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 Specifications 3.3.5.1
| |
| : 3. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 4. Technical Specifications 3.4.4
| |
| : 5. UFSAR 5.2.5
| |
| : 6. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 7. QOA 900-4 A-17, Annuciator Response
| |
| : 8. QGA 100, RPV Control
| |
| : 9. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 10. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 11. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-130 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 Initiating Condition:
| |
| UNPLANNED loss of RPV 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 water level to above the procedurally established lower limit for > 15 minutes.
| |
| OR
| |
| : 2. a. RPV water level cannot be monitoredunknown 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 in RPV inventory Month 20XX QC 3-131 EP-AA-1006 (Revision XX)
| |
| | |
| rl"nd Mac Annoy I~volrln Nl,,Aalr A. i~ ri Ci$i~ Anng~v Fvalnn NImmrI~2r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (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 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.
| |
| 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 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 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-132 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev. 6 CU1
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QGA 100, RPV Control
| |
| : 6. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 7. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 8. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-133 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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, 5, 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 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)
| |
| * RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-134 EP-AA-1006 (Revision XX)
| |
| | |
| Oiiad Cities Annex Exelon NuceaIzr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 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
| |
| : 2. Station Security Plan - Appendix C Month 20XX QC 3-135 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Ou~dCitesRECOGNITION AnexExelon CATEGORY Nuclear HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:
| |
| HOSTILE ACTION within the PROTECTED AREA.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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].
| |
| Month 20XX QC 3-136 EP-AA-1006 (Revision XX)
| |
| | |
| tliinrl r-iflac Annov :valnn kJlrlAa-r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 (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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HS1
| |
| : 3. Station Security Plan - Appendix C Month 20XX QC 3-137 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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.
| |
| Month 20XX QC 3-138 EP-AA-1006 (Revision XX)
| |
| | |
| l:Y*Inn Ninl*_*r Aiisad Citiecs Annex FvaInn N~iJpmrIz RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (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 QCOA 0010-20, Security Event.
| |
| EAL #2 Basis Is 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 HSI.
| |
| Month 20XX QC 3-139 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HA1
| |
| : 2. Station Security Plan - Appendix C
| |
| : 3. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-140 EP-AA-1 006 (Revision XX)
| |
| | |
| Fxelnn NHP.IeRr Oijad !ties Annex FYAlnn Nucler~I RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:
| |
| Confirmed SECURITY CONDITION or threat.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| Basis:
| |
| 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.
| |
| Month 20XX QC 3-141 EP-AA-1006 (Revision XX)
| |
| | |
| l=xnlnn Nucl*nr Quad Cities Annex ExelIn Nucrileazr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HUI (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-101-132.
| |
| 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 QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| 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, HU1
| |
| : 2. SY-AA-101-132, Security Assessment and Response to Unusual Activities
| |
| : 3. Station Security Plan - Appendix C
| |
| : 4. NRC Safeguards Advisory 10/6/01
| |
| : 5. Letter from Mr. B. A. Boger (NRC) to Ms. Lynette Hendricks (NEI) dated 2/4/02
| |
| : 6. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-142 EP-AA-1006 (Revision XX)
| |
| | |
| fl"fkrl difiac Annov IPvlnn hlrlanl*r Au .~ r4 (~i*~
| |
| . L* -- Ann~v ~st Fvalnn
| |
| ~ . ~. . *Muiri~ar U - - *~~*
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition:
| |
| Inability to control a key safety function from outside the Control Room.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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
| |
| " QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| " QOA 0010-05, Plant Operation with the Control Room Inaccessible AND
| |
| : 2. Control of ANY Table H1 key safety function is not reestablished in < 30 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.
| |
| Month 20XX QC 3-143 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)
| |
| Basis (cont):
| |
| 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 30 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-144 EP-AA-1006 (Revision XX)
| |
| | |
| FxA~lnn NnwIArler f3uad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:
| |
| * QCARP 0050-01, SB-i-i Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| * QOA 0010-05, Plant Operation with the Control Room Inaccessible 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-i-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-145 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Exelon Nuclear Nuclear Quad Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition:
| |
| FIRE potentially degrading the level of safety of the plant.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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)
| |
| " Main Control Room Envelope
| |
| * Unit and Shared Emergency Diesel Generator Rooms
| |
| * 4KV Switchgear Area
| |
| " Battery Rooms
| |
| " RHR Service Water Vaults
| |
| " Turbine Building Cable Tunnel
| |
| " Cribhouse 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 Month 20XX QC 3-146 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn Fxelnn Nuclear Nucleazr Ouadr Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-minutes of the initial report, alarm or indication.
| |
| OR
| |
| : 4. A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
| |
| 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.
| |
| Month 20XX QC 3-147 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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.
| |
| 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."
| |
| Month 20XX QC 3-148 EP-AA-1006 (Revision XX)
| |
| | |
| rl.r ar4(ifac~ Annoyv IFYalnn Niir~lanr
| |
| - Fv~Inn
| |
| ~ M.us-102r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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 HA3.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU4
| |
| : 2. UFSAR Section 3.2 Month 20XX QC 3-149 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Nuclear QuadCites AnexExelon RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition:
| |
| Seismic event greater than OBE levels.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| : 1. GGe~trel Roomn personnel fool an actual or potential sE;ismcce AND
| |
| : 2. The occUrrence Of a- eVent is confirmed in a Gesi manner deemed appropriate by the Shift Manager or Emergency Director.
| |
| Basis:
| |
| This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)1 . 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, 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 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.
| |
| Month 20XX QC 3-150 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY 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 Basis This 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 EAL wording 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.
| |
| Month 20XX QC 3-151 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU2
| |
| : 2. QCOA 0010-09, Earthquake Month 20XX QC 3-152 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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.
| |
| Operating Mode Applicability:
| |
| 4-1*2, 3, 4, &ý-D 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.
| |
| TableH3 Areas with Entry Related Mode Applicability A~ea Entry Related Mode
| |
| _ _ _ApplabMDy
| |
| _ _
| |
| ReacEto Bu ilding* Mode 3 and 4 Tue Bu*d~ng* _ _ _ _3
| |
| *Areas required to establish shutdown coo)inRg Table H3 Areas with Entry Related Mode Applicability Area Unit Entry Related Mode Applicability Reactor Building
| |
| * First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 Month 20XX QC 3-153 EP-AA-1006 (Revision XX)
| |
| | |
| fl"ari Mae Annov gI Ir% K-I-F n -&
| |
| hI- r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY AND
| |
| : 2. Entry into the room or area is prohibited or impeded 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).
| |
| 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.
| |
| Month 20XX QC 3-154 EP-AA-1 006 (Revision XX)
| |
| | |
| tliinrl Mina Annov I::Ynlen Kiil,-lar
| |
| ~- * .~
| |
| A ~'4-r;*;~a La~
| |
| * t* Annoy 0 ''* * *
| |
| * fl Fvalrhn. .Mm.r~Io~r
| |
| ~ . ~
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY
| |
| " 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 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 ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA5
| |
| : 2. UFSAR Section 3.2
| |
| : 3. ACIT 660892-20, Station Halon Discharge IDLH Evaluation Month 20XX QC 3-155 EP-AA-1006 (Revision XX)
| |
| | |
| IEx~lon NuclIazr Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 Initiating Condition:
| |
| Hazardous Event Operating Mode Applicability:
| |
| 1,2,3,4,5, 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 River level, as indicated by EITHER:
| |
| : a. High river water level > 594 ft.
| |
| OR
| |
| : b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam # 14 has failed.
| |
| Basis:
| |
| PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| Month 20XX QC 3-156 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
| |
| Month 20XX QC 3-157 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| Basis (cont):
| |
| 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 of power to a SAFETY SYSTEM comrponent as a result of leakage is an event of lesser impact and would be expected to cause small and localized damage. The consequence af thk tuine of event i6 adequaiitel" asese and ird~ressced fin Annn~r'-1nno- With' Technical Specifications.
| |
| 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, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
| |
| This EAL is not intended 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.
| |
| Month 20XX QC 3-158 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| Basis (cont):
| |
| EAL#5 Basis:
| |
| The Design Flood elevation is 594.5 ft. el. (rounded down to 594 ft. el. MSL). This initial design flood elevation is equal to the plant grade of 594.5 ft. el. and any mode of operation is, therefore, possible without additional protective measures.The station design is such that if Lock and Dam No. 14 were to fail, the water level would recede in the intake bay to the point where it would be separated from the river. As the water level recedes in the intake bay, circulating water, service water and fire diesel pumps would become inoperable, leaving only RHRSW and DGCW available to shutdown the units.
| |
| Use of the ultimate heat sink to shutdown the reactors requires the operation of portable diesel pumps with a total capacity of 5100 gpm to reverse the normal flow of makeup water. Makeup water would be provided from the river through the discharge piping and return to the river across the log boom in the intake bay.
| |
| Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, H or C.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU3
| |
| : 2. UFSAR Section 3.2
| |
| : 3. QCTP 0130-11 Internal Flood Protection Program
| |
| : 4. Drawing FL-1 Flood Barriers
| |
| : 5. Quad Cities Nuclear Power Station Unit 1 and 2 Internal Flooding Analysis Note Book, July 1993 Final Draft, prepared by Individual Plant Evaluation Partnership (IPEP)
| |
| Month 20XX QC 3-159 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Nuclear Quad Cities Annex Eeo ula 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, 4,5, 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 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 Month 20XX QC 3-160 EP-AA-1006 (Revision XX)
| |
| | |
| rliinfl 9'ifiaa Annoy I:=v,'lI*n M,,r-la-nr flu .~ r4 ~ Anngw ~vaJr~n kim ~
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Initiating Condition:
| |
| 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, 5, 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.
| |
| 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 a Site Area Emergency.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS7 Month 20XX QC 3-161 EP-AA-1006 (Revision XX)
| |
| | |
| Fxplnn NuJe.l*r Quad Cities Annex FYAIon Nuclerkr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Initiating Condition:
| |
| Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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.
| |
| 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 Month 20XX QC 3-162 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Applicability:
| |
| 1,2,3,4,5, 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:
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU7 Month 20XX QC 3-163 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:
| |
| * > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR 0 > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts
| |
| .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 ISFSls are covered under ICs HU1 and HAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 E-HU1
| |
| : 2. Certificate of Compliance No. 1014 Appendix A Section 5.7 Month 20XX QC 3-164 EP-AA-1006 (Revision XX)
| |
| | |
| l=xpInn NnJP.le_*r Duad Cities Annex Exelnn Nucler~I RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 Initiating Condition:
| |
| Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRem TEDE or 5000 mRem thyroid CDE.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond 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.
| |
| Month 20XX QC 3-33 EP-AA-1006 (Revision XX)
| |
| | |
| FxA~lnn Nuclear Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 (cont)
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG1
| |
| : 2. EP-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 4. QGA 400 Radioactivity Release Control Month 20XX QC 3-34 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| 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.
| |
| Month 20XX QC 3-35 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex QuadCites AnexExellon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RSl (cont)
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AS1
| |
| : 2. EP-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-36 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 Applicability:
| |
| 1,2, 3,4, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
| |
| : a. > 10 mRem TEDE OR
| |
| : b. > 50 mRem CDE Thyroid OR Month 20XX QC 3-37 EP-AA-1006 (Revision XX)
| |
| | |
| Explon Eeo Nuclearula Quad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
| |
| : a. 10 mRem TEDE for 60 minutes of exposure OR
| |
| : b. 50 mRem CDE Thyroid for 60 minutes of exposure OR
| |
| : 4. Field survey results at or beyond the site boundary indicate EITHER:
| |
| : a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for > 60 minutes.
| |
| OR
| |
| : b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Basis:
| |
| This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of 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.
| |
| Escalation of the emergency classification level would be via IC RS1.
| |
| Month 20XX QC 3-38 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelnn Nuclear Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RAI (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA1
| |
| : 2. CY-QC-120-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. QOP 2000-24, Discharging to the River from the River Discharge Tank using the Waste Surge Pump
| |
| : 6. QOP 2000-25, Discharging to the River from the River Discharge Tank using the River Discharge Pump
| |
| : 7. CY-QC-120-729, Liquid Effluent Alarm Setpoints
| |
| : 8. QCOA 1700-02, High Radiation detected on Eberline Radiation Monitoring System
| |
| : 9. QCAN 912-5 C-6, Radwaste High Rad.
| |
| : 10. QCAN 901(2)-3-G-1, Liquid Process Rad. Monitor High Radiation
| |
| : 11. CY-QC-120-735, Main Chimney & Reactor Vent Noble Gas Release Rate Action Levels
| |
| : 12. QCOA 1700-01, Abnormal Chimney Radiation
| |
| : 13. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 14. QGA 400 Radioactivity Release Control Month 20XX QC 3-39 EP-AA-1006 (Revision XX)
| |
| | |
| n"nd rifica Annoy PI=:,len hN"nlar Amm2rI
| |
| -- - - Ci*ia~
| |
| - * ** -- Ann~v g-* ** Fvolnn
| |
| ~st~ *
| |
| * U - ieIa~r
| |
| *
| |
| * NI. * - L4*
| |
| RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 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, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
| |
| : 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for > 60 minutes.
| |
| " Radwaste Effluent Monitor 1/2-1799-01 OR
| |
| " Discharge Permit specified monitor OR
| |
| : 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
| |
| Basis:
| |
| This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
| |
| Month 20XX QC 3-40 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Ou~dCites AnexExellon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| EAL #1 Basis This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
| |
| The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.
| |
| EAL #2 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous effluent pathways.
| |
| EAL #3 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU1
| |
| : 2. CY-QC-1 20-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-41 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, 6, 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.
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG2 Month 20XX QC 3-42 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS2 Initiating Condition:
| |
| Spent fuel pool level at (site-specific Level 3 description).
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 6, D Emergency Action Level (EAL):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, AS2 Month 20XX QC 3-43 EP-AA-1006 (Revision XX)
| |
| | |
| I=xelon Nuclear Ouadr Cities Annex Exelon Nuclea~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS i RADIOLOGICAL EFFLUENTS RA2 Initiating Condition:
| |
| Significant lowering of water level above, or damage to, irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1,2, 3,4, 5, D Emergency Action Level (EAL):
| |
| : 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.
| |
| OR
| |
| : 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R1 Radiation Monitor reading >1000 mRem/hr.
| |
| OR
| |
| : 3. Lowering of spent fuel pool level to (site specific Level 2 value).
| |
| Table R1 Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Monitor
| |
| * 1(2) 1705-16B Fuel Pool Rad Monitor Basis:
| |
| REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.
| |
| IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
| |
| CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-44 EP-AA-1006 (Revision XX)
| |
| | |
| Oiiad Cities~ Annex Exellnn Nuclear I=x*lnn Nuclea~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 (cont)
| |
| Basis (cont):
| |
| 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 an 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 Ror C ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA2
| |
| : 2. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity
| |
| : 3. QCAN 901(2)-3 B-1 Refuel Floor Hi Radiation
| |
| : 4. QCAN 901(2)-3 G-16/H-16 Fuel Pool Channel A/B Hi Radiation
| |
| : 5. QCIS 1700-07 Reactor Building Ventilation and Fuel Pool Radiation Monitoring Calibration and Functional Test Month 20XX QC 3-45 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 Initiating Condition:
| |
| UNPLANNED loss of water level above irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| : 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
| |
| * Refueling Cavity water level < 282 in. (Upper Wide range simulated signal).
| |
| OR
| |
| * Spent Fuel Pool water level < 19 ft. above the fuel (< - 4 ft. indicated level).
| |
| OR
| |
| * Indication or report of a drop in water level in the REFUELING PATHWAY.
| |
| AND
| |
| : b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table RI.
| |
| Table RI Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Mon
| |
| * 1(2) 1705-16B Fuel Pool Rad Mon Month 20XX QC 3-46 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 (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.
| |
| 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 an 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU2
| |
| : 2. QCOP 0201-13 Reactor Level Upper Wide Range Reference Leg Extension Use and Control
| |
| : 3. Technical Specifications 3.7.8 Spent Fuel Storage Pool Water Level
| |
| : 4. Technical Specifications 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel
| |
| : 5. QCAN 901(2)-4 B-24 FUEL POOL STORAGE HI/LO LEVEL
| |
| : 6. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity Month 20XX QC3-47 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, 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 R2 areas:
| |
| Table R2 Areas Requiring Continuous Occupancy
| |
| * Main Control Room (Unit 1 ARM Station #22)
| |
| * Central Alarm Station - (by survey)
| |
| OR
| |
| : 2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:
| |
| Table R3 Areas with Entry Related Mode Applicability Area Unit Entry Related Mode Applicability Reactor Building
| |
| " First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| " First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 Month 20XX QC 3-48 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS 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 R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.
| |
| This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
| |
| Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.
| |
| For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in 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.
| |
| * 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.).
| |
| Month 20XX QC 3-49 EP-AA-1006 (Revision XX)
| |
| | |
| Exnlon Exellon Nuclnnr Nuclenar Quad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA3 (cont)
| |
| Basis (cont):
| |
| " 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. QCOP 1800-1 Operation of ARM Indicator/Trip Units
| |
| : 3. UFSAR Section 3.2
| |
| : 4. General Arrangement Drawings M-5, 6, 8 and 10 Month 20XX QC 3-50 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU3 Initiating Condition:
| |
| Reactor coolant activity greater than Technical Specification allowable limits.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| : 1. Offgas system radiation monitor HI-HI alarm.
| |
| OR
| |
| : 2. Specific coolant activity > 4.0 uCi/gm 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU3
| |
| : 2. Technical Specifications 3.4.6
| |
| : 3. Technical Specifications 3.7.6
| |
| : 4. QCAN 901(2)-3 C-2 OFF GAS HIGH-HIGH RADIATION Month 20XX QC 3-51 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:
| |
| Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the General Emergency classification level each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-52 EP-AA-1006 (Revision XX)
| |
| | |
| 11iind Mfip4m Annoy FIylnn mnmirlar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:
| |
| Loss or Potential Loss of ANY two barriers.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the Site Area Emergency classification level, each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-53 EP-AA-1 006 (Revision XX)
| |
| | |
| (3"nri Mtiac. AnnpX Pvalnn khirlanar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:
| |
| ANY Loss or ANY Potential Loss of EITHER Fuel Clad or RCS.
| |
| Operating Mode Applicability:
| |
| 1,2,3 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 Month 20XX QC 3-54 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Nuclear Quad Cities Annex Exelon NuclIear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:
| |
| RCS Activity Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Coolant activity > 300 uCi/gm Dose Equivalent 1-131.
| |
| Basis:
| |
| This threshold indicates that RCS radioactivity concentration is greater than 300 PCi/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
| |
| : 2. NF-AA-430, Failed Fuel Action Plan Month 20XX QC 3-55 EP-AA-1 006 (Revision XX)
| |
| | |
| tlimri riflac Annoy =,v,-lInn I*1" t-I--r flu .~ ii -...--
| |
| ---- Citia~ m.
| |
| Anngw ~vjt~iIw~n
| |
| ~ Mi- ar~I~r I ~S I I EU - l~~I RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:
| |
| RPV Water Level 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 > -142 inches (TAF)
| |
| OR
| |
| : 3. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.
| |
| This is identified in the BWROG EOPs/SAMGs 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 Month 20XX QC 3-56 EP-AA-1 006 (Revision XX)
| |
| | |
| Oiiad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)
| |
| Basis (cont):
| |
| RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
| |
| The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-57 EP-AA-1006 (Revision XX)
| |
| | |
| Explon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Drywell radiation monitor reading > 6.65 E+02 R/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 300 [tCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
| |
| The radiation monitor reading in this threshold is higher than that specified for RCS Barrier RC5 Loss 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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. Core Damage Assessment Methodology (CDAM)
| |
| Month 20XX QC 3-58 EP-AA-1006 (Revision XX)
| |
| | |
| tl"nrl ifine Annoy Pvalnn Pd"d-lanr A. .~rI Citiac Ann~v ~vaInn Mmme-Ia2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-59 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. RPV water level cannot be restored and maintained > -142 inches (TAF)
| |
| OR
| |
| : 2. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.
| |
| The RPV water level threshold is the same as Fuel Clad 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 Month 20XX QC 3-60 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)
| |
| Basis (cont):
| |
| fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 500-4 RPV Flooding Month 20XX QC 3-61 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:
| |
| Primary Containment Pressure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell pressure >2.5 psig.
| |
| AND
| |
| : 2. Drywell pressure rise is due to RCS leakage.
| |
| Basis:
| |
| The > 2.5 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.
| |
| The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.
| |
| The release of mass from the RCS due to the as-designed/expected operation of any relief valve does not warrant an emergency classification.
| |
| A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| There is no Potential Loss threshold associated with Primary Containment Pressure.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 200 Primary Containment Control Month 20XX QC 3-62 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:
| |
| RCS Leak Rate Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC 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 > QGA 300 Maximum Normal operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300 Maximum Normal 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 on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.
| |
| 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.
| |
| Month 20XX QC 3-63 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis (cont):
| |
| 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 safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the Torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
| |
| Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
| |
| A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
| |
| The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.
| |
| In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area. radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.
| |
| An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold
| |
| #1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
| |
| Month 20XX QC 3-64 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex OuadCitesRECOGNITION AnexExellon CATEGORY Nuclear FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. M-13 Main steam piping
| |
| : 3. UFSAR 5.2.5
| |
| : 4. QCOA 0201-01 Rev 016, Increasing Drywell Pressure
| |
| : 5. QOA 900-4 A-17 900-4 A-17 Annunciator
| |
| : 6. QCOS 1600-07 Reactor Coolant Leakage In The Drywell
| |
| : 7. QGA 300 Secondary Containment Control Month 20XX QC 3-65 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:
| |
| Primary Containment radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell radiation monitor reading > 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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. CaIc. EP-EAL-0611 Month 20XX QC 3-66 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:
| |
| 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.
| |
| 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 Month 20XX QC 3-67 EP-AA-1006 (Revision XX)
| |
| | |
| Fxolnn N Hc.lo*r Quad Cities Annex Fxelnn Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-68 EP-AA-1006 (Revision XX)
| |
| | |
| i=x*lnn NHcle.nr Oun~d Cities Annex Fvnlnn NuclerIir RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:
| |
| Primary Containment 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 > 56 psig and rising.
| |
| OR
| |
| : 4. a. Drywell or torus hydrogen concentration > 6%.
| |
| AND
| |
| : b. Drywell or torus oxygen concentration > 5%.
| |
| OR
| |
| : 5. Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| Loss Threshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
| |
| These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.
| |
| Month 20XX QC 3-69 EP-AA-1006 (Revision XX)
| |
| | |
| f-)"nd rifiac Annov I=Y,-I~n I~el ln-r A *1214 ~ Anntw ~v~Ir~n MI.ria2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)
| |
| Basis (cont):
| |
| Potential Loss Threshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.
| |
| Potential Loss 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. UFSAR Fig. 6.2-16a
| |
| : 3. UFSAR Section 15.6
| |
| : 4. UFSAR 6.2.1.1
| |
| : 5. QGA 200, Primary Containment Control
| |
| : 6. Quad Cities PSTG Section 5, Primary Containment Control Month 20XX QC 3-70 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS
| |
| : 1. Drywell radiation monitor reading > 1.55 E+03 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)
| |
| Month 20XX QC 3-71 EP-AA-1006 (Revision XX)
| |
| | |
| A. iiinf r~ifin Annav P~vlf~n Niu,,nlar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:
| |
| Primary Containment Isolation Failure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 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 > QGA 300, Maximum Safe operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300, Maximum Safe operating levels.
| |
| 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, HPCI or RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.
| |
| Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.
| |
| Month 20XX QC 3-72 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Month 20XX QC 3-73 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)
| |
| Basis (cont):
| |
| Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.
| |
| In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.
| |
| 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. QGA 200 Primary Containment Control
| |
| : 3. QGA 200-5 Hydrogen Control
| |
| : 4. QCOP 1600-13 Post-Accident Venting of the Primary Containment
| |
| : 5. QGA 300 Secondary Containment Control Month 20XX QC 3-74 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-75 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex OundCites AnexExelon Nuclear 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. EITHER of the following:
| |
| : a. Restoration of at least one unit ECCS bus in < 1 hour is not likely.
| |
| OR
| |
| : b. RPV water level cannot be restored and maintained > -166 inches.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.
| |
| This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat 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 FGI. This will allow additional time for implementation of offsite protective actions.
| |
| Month 20XX QC 3-76 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)
| |
| Basis (cont):
| |
| Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
| |
| The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
| |
| The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. QGA 100 RPV Control Month 20XX QC 3-77 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lnn Nu*le_*r Quad Cities Annex Exelnn Nuclea~r 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECOS. 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 or MG1.
| |
| Month 20XX QC 3-78 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lnn NH*lpar Ouadc Cities Annex Exelon Nucler~I RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-79 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit Auxiliary Transformer TR-11 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers AND
| |
| : 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.
| |
| 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 MUl.
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| * A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
| |
| Month 20XX QC 3-80 EP-AA-1006 (Revision XX)
| |
| | |
| Fxplnn NHc.le_*r ODiad C~ities Annex FvAlnn Nucleasr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)
| |
| Basis (cont):
| |
| * A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-81 EP-AA-1006 (Revision XX)
| |
| | |
| Oijad Cities Annex ExeIon Nuclear I=x*lon Nuc~lear 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:
| |
| 1,2,3 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 unit ECCS busses 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. (e.g. unit cross-tie breakers)
| |
| Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
| |
| Escalation of the emergency classification level would be via IC MAI.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SUW
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout Month 20XX QC 3-82 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex RECOGNITION CATEGORY Exelon Nuclear SYSTEM MALFUNCTIONS 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.
| |
| AND
| |
| : 3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.
| |
| AND
| |
| : 4. ALL AC and Vital DC power sources have been lost for >15 minutes.
| |
| 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 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 EALs are met.
| |
| Month 20XX QC 3-83 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lnn NHc.le_*r Quad Cities Annex Exellnn Nulairhr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG8
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. Technical Specifications 3.8.4 and B3.8.4
| |
| : 10. UFSAR Section 8.3.2
| |
| : 11. QOP 6900-02 125 VDC Electrical System
| |
| : 12. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-84 EP-AA-1006 (Revision XX)
| |
| | |
| I=xAInn N=m.lp_*r Oiiad Cities Annex FYAIon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2 Initiating Condition:
| |
| Loss of all vital DC power for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| Voltage is < 105 VDC on 125 VDC battery busses #1 and #2 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS8
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-85 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 > 5%.
| |
| AND
| |
| : 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
| |
| AND
| |
| : 3. EITHER of the following conditions exist:
| |
| * RPV water level cannot be restored and maintained > -166 inches.
| |
| OR
| |
| * Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| 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.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Escalation of the emergency classification level would be via IC RG1 or FGI.
| |
| Month 20XX QC 3-86 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 200 Primary Containment Control Month 20XX QC 3-87 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2 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 > 5%.
| |
| AND
| |
| : 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.
| |
| Basis:
| |
| This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.
| |
| Month 20XX QC 3-88 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| Month 20XX QC 3-89 EP-AA-1 006 (Revision XX)
| |
| | |
| Duad Cities Annex Exellon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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
| |
| > 5%.
| |
| 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
| |
| > 5%.
| |
| 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 control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
| |
| Month 20XX QC 3-90 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.
| |
| The plant response to the failure of an automatic or manual reactor 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 control 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 FA1. 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.
| |
| Month 20XX QC 3-91 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exallon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. Technical Specifications Table 3.3.1.1-1
| |
| : 5. Technical Specification 3.3.1.3
| |
| : 6. Technical Specification Bases 3.3.1.1 and 3.3.1.3 Month 20XX QC 3-92 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| * Reactor Power
| |
| * RPV Water Level
| |
| * RPV Pressure
| |
| * Drywell Pressure
| |
| " Torus Level
| |
| " Torus Temperature AND
| |
| : 2. ANY Table M2 transient in progress.
| |
| Table M2 Significant Transients
| |
| " Turbine Trip
| |
| * Reactor Scram
| |
| * ECCS Activation
| |
| * Recirc. Runback > 25% Reactor Power Change
| |
| " Thermal Power oscillations > 10% Reactor Power Change Month 20XX QC 3-93 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear 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 EGGS. 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 Month 20XX QC 3-94 EP-AA-1006 (Revision XX)
| |
| | |
| f_)"nri rifina Annov I=Yalnn N,,r~lanr A.
| |
| - - u~
| |
| - -ri (~iti~
| |
| - . S. -- ~Annoy U*
| |
| * S FvciInn
| |
| - MImr102r U - ~*
| |
| RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| * Reactor Power
| |
| * RPV Water Level
| |
| * RPV Pressure
| |
| * Drywell Pressure
| |
| * Torus Level
| |
| * Torus Temperature Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
| |
| As used in this EAL, an "inability to monitor" means that values for 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.
| |
| Month 20XX QC 3-95 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex I=x*lon Nucler~I Nuclear Ouad itiesAnnexExeIon 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 Month 20XX QC 3-96 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 Initiating Condition:
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-97 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lon NucIp_*r Ouaid C~ities~Annex ExeInn Nucrlear 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.
| |
| This EAL #2a 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.
| |
| This 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 is not 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 SA9 Month 20XX QC 3-98 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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.
| |
| 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 Month 20XX QC 3-99 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)
| |
| Basis (cont):
| |
| 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 Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| 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. QCOS 1600-07 Reactor Coolant Leakage in the Drywell
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QCOA 0201-01 Increasing Drywell Pressure Month 20XX QC 3-100 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelnn Nuclear I=x*lon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Initiating Condition:
| |
| Loss of all On-site or Off-site communications capabilities.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency 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 Radio X Plant Page X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite 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.).
| |
| Month 20XX QC 3-101 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis This EAL addresses a total loss of the communications methods used in support of routine plant operations.
| |
| EAL #2 Basis This EAL 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) Form.
| |
| EAL #3 Basis This EAL 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-102 EP-AA-1006 (Revision XX)
| |
| | |
| IFxAlnn NurlIa~r Duad Cities Annex Exelon Nuclear 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:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. Loss of ALL offsite AC power to unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| 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.
| |
| Month 20XX QC 3-103 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities. Annex OuadCites AnexExellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA2
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "ACTURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-104 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS cul Initiating Condition:
| |
| Loss of all but one AC power source to emergency buses for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. AC power capability to unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit auxiliary transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers 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.
| |
| Month 20XX QC 3-105 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU1 (cont)
| |
| Basis (cont):
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| " 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-106 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:
| |
| Hazardous event affecting SAFETY SYSTEM required for the current operating mode.
| |
| Operating Mode Applicability:
| |
| 4,5 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.
| |
| 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.
| |
| Month 20XX QC 3-107 EP-AA-1006 (Revision XX)
| |
| | |
| A, .rl riCiiac Annoy Pvnlrn Khid-In!mr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (cont)
| |
| Basis (cont):
| |
| 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, 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA6 Month 20XX QC 3-108 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:
| |
| Loss of Vital DC power 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.
| |
| Voltage is < 105 VDC on required 125 VDC battery busses #1 and #2 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 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.
| |
| Month 20XX QC 3-109 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CU4
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-110 EP-AA-1006 (Revision XX)
| |
| | |
| Mind Mae Annov I,-Invlrn M, "Alr fl112r4
| |
| - - 4 (~i*iaQ
| |
| -
| |
| * t* S - Ann~v S~* U* *fl ~vaIr~n fl * * * *Mi U - ir~Ia~r
| |
| * 4*
| |
| RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:
| |
| Loss of all onsite or offsite communications capabilities.
| |
| Operating Mode Applicability:
| |
| 4,5, 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 Radio X Plant Page X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite communications capabilities.
| |
| While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Outside 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.
| |
| Month 20XX QC 3-111 EP-AA-1006 (Revision XX)
| |
| | |
| tl"nri Pifipc Annov I=Yaltnn NIuid-lar 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-112 EP-AA-1006 (Revision XX)
| |
| | |
| tliumrl ritioa Annoy F=volnn N"Ii-lanr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 Initiating Condition:
| |
| Inability to maintain the plant in cold shutdown.
| |
| 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.
| |
| " 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*
| |
| Not Intact Established 20 minutes*
| |
| 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.
| |
| 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.
| |
| 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.).
| |
| Month 20XX QC 3-113 EP-AA-1006 (Revision XX)
| |
| | |
| Mind Mac Annov I:::a--In N"Aal--r A m.2,4 Ci$i~ Annav Fv~Inn NiirIa2r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Month 20XX QC 3-114 EP-AA-1006 (Revision XX)
| |
| | |
| Otiad Cities Annex Exellon Nuclear I=x*lon Nucla~r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. OU-AA-103 Shutdown Safety
| |
| : 6. QCOA 1000-02 Loss of Shutdown Cooling
| |
| : 7. QGA 100 RPV Control
| |
| : 8. QGA 100 RPV Control Detail A
| |
| : 9. QCGP 1-1 Normal Unit Startup
| |
| : 10. QCIS 0600-01 Unit One Division 1 Reactor Pressure 0 to 1200 psig Indication Calibration Month 20XX QC 3-115 EP-AA-1006 (Revision XX)
| |
| | |
| I:x*lnn Nuclear Oiiad Cities Annex Exelnn Nuclea~r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 Initiating Condition:
| |
| UNPLANNED rise in RCS temperature 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.
| |
| * 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.).
| |
| A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
| |
| Month 20XX QC 3-116 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 (cont)
| |
| Basis (cont):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, CU3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. QGA 100, RPV Control
| |
| : 4. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 5. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 6. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge
| |
| : 7. QCOA 1000-02 Loss of Shutdown Cooling Month 20XX QC 3-117 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Initiating Condition:
| |
| Loss of RPV inventory affecting fuel clad integrity with containment challenged.
| |
| 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 a. RPV water level < -142 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
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-118 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| Table C4 Containment Challenge Indications
| |
| " Primary Containment Hydrogen Concentration > 6% and Oxygen > 5%
| |
| * UNPLANNED rise in containment pressure
| |
| * CONTAINMENT CLOSURE not established*
| |
| * ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.
| |
| * if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.
| |
| 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.
| |
| Month 20XX QC 3-119 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis (cont):
| |
| 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-120 EP-AA-1006 (Revision XX)
| |
| | |
| l:y*Inn N.*lp_;*r Ouasd Cities Annex FvAllnn Nula~r~I RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CG1
| |
| : 2. QGA 100, RPV Control
| |
| : 3. Technical Specifications 3.3.1
| |
| : 4. Technical Specifications 3.6.1.1
| |
| : 5. Technical Specifications 3.6.4.1
| |
| : 6. QGA-200-5, Hydrogen Control
| |
| : 7. UFSAR 6.2.1.1
| |
| : 8. QGA 300 Secondary Containment Control
| |
| : 9. EP-EAL-0501, Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Month 20XX QC 3-121 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 Initiating Condition:
| |
| Loss of RPV inventory affecting core decay heat removal capability.
| |
| 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. With CONTAINMENT CLOSURE not established, RPV water level
| |
| <- 65 inches OR
| |
| : 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 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
| |
| * Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-122 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 (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.
| |
| 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.b and 2.b 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 tank levels. Sump and/or tank 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 RG1.
| |
| Month 20XX QC 3-123 EP-AA-1006 (Revision XX)
| |
| | |
| E~In~ln Exellon Nclea~r Nuclear Ouad ities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CS1
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. QGA 100, RPV Control
| |
| : 6. Technical Specifications 3.3.1
| |
| : 7. Technical Specifications Table 3.3.3.1-1
| |
| : 8. Technical Specifications 3.3.5.1
| |
| : 9. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 10. Technical Specifications 3.4.4
| |
| : 11. UFSAR 5.2.5
| |
| : 12. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 13. QOA 900-4 A-17, Annuciator Response Month 20XX QC 3-124 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:
| |
| Loss of RPV 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 < - 59 inches.
| |
| OR
| |
| : 2. a. RPV water level cannot be monitored for > 15 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 in 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 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.
| |
| Month 20XX QC 3-125 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis A lowering of water level below -59 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 decrease in water level will lead to core uncovery.
| |
| 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). An 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 tank levels. Sump and/or tank 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 Specifications 3.3.5.1
| |
| : 3. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 4. Technical Specifications 3.4.4
| |
| : 5. UFSAR 5.2.5
| |
| : 6. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 7. QOA 900-4 A-17, Annuciator Response
| |
| : 8. QGA 100, RPV Control
| |
| : 9. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 10. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 11. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-126 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex QuadCites AnexExelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 Initiating Condition:
| |
| UNPLANNED loss of RPV 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 water level to 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.
| |
| 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 in RPV inventory Month 20XX QC 3-127 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (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 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.
| |
| 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 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 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-128 EP-AA-1 006 (Revision XX)
| |
| | |
| Exallon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev. 6 CUI
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QGA 100, RPV Control
| |
| : 6. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 7. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 8. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-129 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, 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 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)
| |
| * RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-130 EP-AA-1 006 (Revision XX)
| |
| | |
| t'J".qd riflac Anna-z I=v,',Inn M,,rl-lar RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 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 ContingencyPlan [and Independent Spent Fuel Storage Installation Security Program].
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev. 6 HG1
| |
| : 2. Station Security Plan - Appendix C Month 20XX QC 3-131 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:
| |
| HOSTILE ACTION within the PROTECTED AREA.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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].
| |
| Month 20XX QC 3-132 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 (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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HS1
| |
| : 3. Station Security Plan - Appendix C Month 20XX QC 3-133 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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.
| |
| Month 20XX QC 3-134 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (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 QCOA 0010-20, Security Event.
| |
| EAL #2 Basis Is 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.
| |
| Month 20XX QC 3-135 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HA1
| |
| : 2. Station Security Plan - Appendix C
| |
| : 3. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-136 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:
| |
| Confirmed SECURITY CONDITION or threat.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| Basis:
| |
| 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.
| |
| Month 20XX QC 3-137 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HUl (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-101-132.
| |
| 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 QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| 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, HU1
| |
| : 2. SY-AA-1 01-132, Security Assessment and Response to Unusual Activities
| |
| : 3. Station Security Plan - Appendix C
| |
| : 4. NRC Safeguards Advisory 10/6/01
| |
| : 5. Letter from Mr. B. A. Boger (NRC) to Ms. Lynette Hendricks (NEI) dated 2/4/02
| |
| : 6. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-138 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition:
| |
| Inability to control a key safety function from outside the Control Room.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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
| |
| * QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| * QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR QOA 0010-05, Plant Operation with the Control Room Inaccessible AND
| |
| : 2. Control of ANY Table H1 key safety function is not reestablished in < 30 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.
| |
| Month 20XX QC 3-139 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)
| |
| Basis (cont):
| |
| 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 30 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-i-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-140 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:
| |
| * QCARP 0050-01, SB-i-i Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| " QOA 0010-05, Plant Operation with the Control Room Inaccessible 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-141 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition:
| |
| FIRE potentially degrading the level of safety of the plant.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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)
| |
| " Main Control Room Envelope
| |
| " Unit and Shared Emergency Diesel Generator Rooms
| |
| * 4KV Switchgear Area
| |
| * Battery Rooms
| |
| " RHR Service Water Vaults
| |
| " Turbine Building Cable Tunnel
| |
| " Cribhouse OR
| |
| : 2. a. Receipt of a sinale 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 Month 20XX QC 3-142 EP-AA-1006 (Revision XX)
| |
| | |
| I=Y*lnn Nirl*r 0.32,1 ritip Annpw FvalInn N"A~canr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-minutes of the initial report, alarm or indication.
| |
| OR
| |
| : 4. A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
| |
| 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.
| |
| Month 20XX QC 3-143 EP-AA-1 006 (Revision XX)
| |
| | |
| Otiad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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.
| |
| 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."
| |
| Month 20XX QC 3-144 EP-AA-1 006 (Revision XX)
| |
| | |
| ('1"nei Mae Annav Pvnlrn Kl"Anmr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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 HA3.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU4
| |
| : 2. UFSAR Section 3.2 Month 20XX QC 3-145 EP-AA-1006 (Revision XX)
| |
| | |
| (_)"nri rifine Annav Pvalnn kl"rlanr A I .2,4 Citia~ Ann~w Fvoinn MIIrI~h2r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition:
| |
| Seismic event greater than OBE levels.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| Basis:
| |
| This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)1 . 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, 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.
| |
| 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 public 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.
| |
| Month 20XX QC 3-146 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon NucerIAm RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)
| |
| Basis (cont):
| |
| 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 EAL wording 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU2
| |
| : 2. QCOA 0010-09, Earthquake
| |
| : 3. US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions.
| |
| Month 20XX QC 3-147 EP-AA-1006 (Revision XX)
| |
| | |
| ir4 Min ifac Anncov PvI::lnn Nelalr 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.
| |
| 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 Unit Entry Related Mode Applicability Reactor Building
| |
| " First Floor North Wall 1
| |
| * Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| " Second Floor South Wall 2 High Pressure Heater Bay 1&2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 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.
| |
| Month 20XX QC 3-148 EP-AA-1006 (Revision XX)
| |
| | |
| I=Y*_lnn NuJi*.lm*ar 0 Ilad C~ities Annex Fvazimnn NKwIAria RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)
| |
| Basis (cont):
| |
| 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 ifthis 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).
| |
| 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.
| |
| Month 20XX QC 3-149 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)
| |
| Basis (cont):
| |
| 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 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 ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA5
| |
| : 2. UFSAR Section 3.2
| |
| : 3. ACIT 660892-20, Station Halon Discharge IDLH Evaluation Month 20XX QC 3-150 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 Initiating Condition:
| |
| Hazardous Event Operating Mode Applicability:
| |
| 1,2,3,4,5, 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 River level, as indicated by EITHER:
| |
| : a. High river water level > 594 ft.
| |
| OR
| |
| : b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam # 14 has failed.
| |
| Basis:
| |
| PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| Month 20XX QC 3-151 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| 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.
| |
| 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, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
| |
| This EAL is not intended 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.
| |
| Month 20XX QC 3-152 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| Basis (cont):
| |
| EAL#5 Basis:
| |
| The Design Flood elevation is 594.5 ft. el. (rounded down to 594 ft. el. MSL). This initial design flood elevation is equal to the plant grade of 594.5 ft. el. and any mode of operation is, therefore, possible without additional protective measures.The station design is such that if Lock and Dam No. 14 were to fail, the water level would recede in the intake bay to the point where it would be separated from the river. As the water level recedes in the intake bay, circulating water, service water and fire diesel pumps would become inoperable, leaving only RHRSW and DGCW available to shutdown the units.
| |
| Use of the ultimate heat sink to shutdown the reactors requires the operation of portable diesel pumps with a total capacity of 5100 gpm to reverse the normal flow of makeup water. Makeup water would be provided from the river through the discharge piping and return to the river across the log boom in the intake bay.
| |
| Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, H or C.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU3
| |
| : 2. UFSAR Section 3.2
| |
| : 3. QCTP 0130-11 Internal Flood Protection Program
| |
| : 4. Drawing FL-1 Flood Barriers
| |
| : 5. Quad Cities Nuclear Power Station Unit 1 and 2 Internal Flooding Analysis Note Book, July 1993 Final Draft, prepared by Individual Plant Evaluation Partnership (IPEP)
| |
| Month 20XX QC 3-153 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, 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 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 Month 20XX QC 3-154 EP-AA-1006 (Revision XX)
| |
| | |
| rl"nrl Mac Annov I=vnlrn kl"Al--r fl.m~mrI
| |
| ~- S.- Ci$i~~
| |
| W ** -- Annoy C-.. * * . ~efl ~voInn
| |
| ~ . ~. . MuurI~~r I. - ~
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Initiating Condition:
| |
| Other conditions exist which in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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 a Site Area Emergency.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS7 Month 20XX QC 3-155 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn N ilrulI_*r 011ad Nties Annex Fxelon Nnelpar RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Initiating Condition:
| |
| Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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 Month 20XX QC 3-156 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Applicability:
| |
| 1,2, 3, 4, 5, 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:
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU7 Month 20XX QC 3-157 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*_lnn NH*I*r (3.q~d Cities~Annex FYAlnn Nucler~I RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:
| |
| 0 > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR 0 > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts
| |
| .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 ISFSs are covered under ICs HU1 and HAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 E-HU1
| |
| : 2. Certificate of Compliance No. 1014 Appendix A Section 5.7 Month 20XX QC 3-158 EP-AA-1006 (Revision XX)}}
| |
| | |
| =Text=
| |
| {{#Wiki_filter:ATTACHMENT 10 REVISED RADIOLOGICAL EMERGENCY PLAN ANNEX INFORMATION FOR QUAD CITIES NUCLEAR POWER STATION EP-AA-1 006 Enclosures 6 Enclosure 10A - Revised EAL Comparison Matrix Document 0 Enclosure 10B - Revised EAL Red-Line Basis Documents 0 Enclosure 10C - Revised EAL Basis Documents
| |
| | |
| Justification 99-01 Proposed EAL I 99-01 Rev NEINEI Rev 66 Proposed EAL Justification Initiating Condition - GENERAL EMERGENCY AGI Initiating Condition:
| |
| RG1 H No Change Difference Deviation Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem Release of gaseous radioactivity resulting in offsite dose greater than 1,000 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 5,000 mrem thyroid CDE. mRem TEDE or 5,000 mRem thyroid CDE. classification.
| |
| Operating Mode Applicability: All Operating Mode Applicability:
| |
| : 2) Added the following to bullet #3 "Classification based on effluent monitor 1,2,3,4,5,D readings assumes that a release path to the environment is established." In order Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): 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 Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to
| |
| * The Emergency Director should declare the General Emergency promptly isolate the release path, then the effluent monitor reading is no longer valid for upon determining that the applicable time has been exceeded, or will likely be classification purposes." This allows for more timely classification since all the
| |
| * The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded.
| |
| that the applicable time has been exceeded, or will likely be exceeded. readily available on the 11x17 procedure matrix used by the SM.
| |
| * If an ongoing release is detected and the release start time is unknown,
| |
| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| assume that the release duration has exceeded 15 minutes.
| |
| I the effluent flow past an effluent monitor is known to have stopped due to if actions to isolate the release path, then the effluent monitor reading is no
| |
| * 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 longer valid for classification purposes.
| |
| known to have stopped due to actions to isolate the release path, then the
| |
| * The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.
| |
| used for emergency classification assessments until the results from a dose The pre-calculated effluent monitor values presented in EAL #1 should be assessment using actual meteorology are available.
| |
| used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
| |
| : 1. Reading on any of the following radiation monitors greater than the reading shown for 15 minutes or longer: 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCilsec for > 15 minutes (as determined by Control Room Panels or PPDS (site specific monitor list and threshold values) - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the
| |
| : 2. Dose assessment actual meteorology indicates doses greater than 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose site boundary of EITHER:
| |
| receptor point) a. > 1000mRemTEDE OR
| |
| : 3. Field survey results indicate EITHER of the following at or beyond (site b. > 5000 mRem CDE Thyroid specific dose receptor point):
| |
| OR
| |
| * Closed window dose rates greater than 1000 mR/hr expected to 3. Field survey results at or beyond the site boundary indicate EITHER:
| |
| continue for 60 minutes or longer.
| |
| : a. Gamma (closed window) dose rates >1000 mR/hr are expected to
| |
| * Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.
| |
| 5000 mrem for one hour of inhalation.
| |
| OR
| |
| : b. Analyses of field survey samples indicate > 5000 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Page 1 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification Initiating Condition - SITE AREA EMERGENCY AS1 Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem Initiating Condition:
| |
| RS1 E No Change E Difference 1 Deviation Release of gaseous radioactivity resulting in offsite dose greater than 100 mRem 1) Listed site-specific monitors and Threshold values to ensure timely TEDE or 500 mrem thyroid CDE. TEDE or 500 mRem thyroid CDE. classification.
| |
| Operating Mode Applicability: All Operating Mode Applicability:
| |
| : 2) Added the following to bullet #3 " Classification based on effluent monitor 1,2,3,4,5,D readings assumes that a release path to the environment is established." In order Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): 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 Notes: Notes: effluent flow past an effluent monitor is known to have stopped due to actions to
| |
| * The Emergency Director should declare the Site Area Emergency promptly isolate the release path, then the effluent monitor reading is no longer valid for upon determining that the applicable time has been exceeded, or will likely be classification purposes." This allows for more timely classification since all the exceeded.
| |
| * The Emergency Director should declare the event promptly upon determining basis information pertaining to Note bullet 3 will be contained in the IC and therefor that the applicable time has been exceeded, or will likely be exceeded. readily available on the 1 x17 procedure matrix used by the SM.
| |
| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| * 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
| |
| * Classification based on effluent monitor readings assumes that a release path longer valid for classification purposes. 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
| |
| * The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes.
| |
| used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. 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:
| |
| : 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCl/sec for> 15 minutes (as determined by Control Room Panels or PPDS -
| |
| (site specific monitor list and threshold values) Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment actual meteorology indicates doses greater than 1000 2. Dose assessment using actual meteorology indicates doses at or beyond the mrem TEDE or 5000 mrem thyroid CDE at or beyond (site specific dose site boundary of EITHER:
| |
| receptor point) a. >100 mRem TEDE OR
| |
| : 3. Field survey results indicate EITHER of the following at or beyond (site b. > 500 mRem CDE Thyroid specific dose receptor point):
| |
| OR
| |
| . Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer. 3. Field survey results at or beyond the site boundary indicate EITHER:
| |
| : a. Gamma (closed window) dose rates >100 mR/hr are expected to
| |
| . Analysis of field survey samples indicate thyroid CDE greater than continue for > 60 minutes.
| |
| 500 mrem for one hour of inhalation.
| |
| OR
| |
| : b. Analyses of field survey samples indicate > 500 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Page 2 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification AA1 RAI Initiating Condition - ALERT Initiating Condition:
| |
| D No Change E Difference 1-1 Deviation Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 1) Listed site-specific monitors and Threshold values to ensure timely classification.
| |
| mrem TEDE or 50 mrem thyroid CDE. mrem TEDE or 50 mrem thyroid CDE.
| |
| Operating Mode Applicability: All Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1, 2, 3,4, 5, D to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): readings assumes that a release path to the environment is established, If the effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for
| |
| * The Emergency Director should declare the Alert promptly upon classification purposes." This allows for more timely classification since all the basis determining that the applicable time has been exceeded, or will likely be information pertaining to Note bullet 3 will be contained in the IC and therefor
| |
| * The Emergency Director should declare the event promptly upon determining readily available on the 1 lx1 7 procedure matrix used by the SM.
| |
| exceeded. that the applicable time has been exceeded, or will likely be exceeded.
| |
| * If an ongoing release is detected and the release start time is unknown,
| |
| * If an ongoing release is detected and the release start time is unknown, assume 3) A calculation was performed, in accordance with (lAW) guidance provided in NEI assume that the release duration has exceeded 15 minutes. that the release duration has exceeded 15 minutes. 99-01 revision 6 EAL AA1, to determine the effluent monitor response for a
| |
| * If the effluent flow past an effluent monitor is known to have stopped due radioactive liquid release via the normal site release pathway. The liquid release Classification based on effluent monitor readings assumes that a release path would contain activity equivalent to provide lmrem TEDE or 50mrem thyroid CDE to actions to isolate the release path, then the effluent monitor reading is to the environment is established. If the effluent flow past an effluent monitor is at the site boundary. The calculation determined the effluent monitor response no longer valid for classification purposes. known to have stopped due to actions to isolate the release path, then the would be >110% of the instrument maximum range and as such, lAW NEI 99-01
| |
| " The pre-calculated effluent monitor values presented in EAL #1 should be effluent monitor reading is no longer valid for classification purposes. Rev 6 guidance, was not included in this EAL.
| |
| used for emergency classification assessments until the results from a The pre-calculated effluent monitor values presented in EAL #1 should be used dose assessment using actual meteorology are available. for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
| |
| : 1. Reading on any of the following radiation monitors greater than the
| |
| : 1. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 reading shown for 15 minutes or longer: uCilsec for > 15 minutes (as determined by Control Room Panels or PPDS
| |
| - Total Noble Gas Release Rate).
| |
| OR (site-specific monitor list and threshold values)
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
| |
| : 2. Dose assessment actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose a. > 10 mRem receptor point) TEDE
| |
| : 3. Analysis of a liquid effluent sample indicates a concentration or release OR rate that would result in doses greater than 10 mrem TEDE or 50 mrem b. > 50 mRem CDE Thyroid thyroid CDE at or beyond (site-specific dose receptor point) for one hour of exposure. OR
| |
| : 3. Analysis of a liquid effluent sample indicates a concentration or release rate
| |
| : 4. Field survey results indicate EITHER of the following at or beyond (site that would result in doses greater than EITHER of the following at or beyond specific dose receptor point): the site boundary
| |
| * Closed window dose rates greater than 10 mR/hr expected to a. 10 mRem TEDE for 60 minutes of exposure continue for 60 minutes or longer.
| |
| OR
| |
| * Analysis of field survey samples indicate thyroid CDE greater than
| |
| : b. 50 mRem CDE Thyroid for 60 minutes of exposure 50 mrem for one hour of inhalation.
| |
| OR
| |
| : 4. Field survey results at or beyond the site boundary indicate EITHER:
| |
| : a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for k 60 minutes.
| |
| OR
| |
| : b. Analyses of field survey sam ples in dicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Page 3 of66
| |
| | |
| NEI 99-01 Rev 6 ] Proposed EAL Justification Initiating Condition - UNUSUAL EVENT AU1 Initiating Condition:
| |
| RUI D No Change E Difference 1 Deviation Release of gaseous or liquid radioactivity greater than 2 times the (site-specific Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 1) Listed site-specific monitors and Threshold values to ensure timely classification.
| |
| effluent release controlling document) limits for 60 minutes or longer 60 minutes or longer.
| |
| Operating Mode Applicability: All Operating Mode Applicability: 2) Added the following to bullet #3" Classification based on effluent monitor readings assumes that a release path to the environment is established." In order 1, 2, 3,4,5, D to delete the following from the basis "Classification based on effluent monitor Example Emergency Action Levels: (1 or 2 or 3 Emergency Action Level (EAL): readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to Note: Notes: isolate the release path, then the effluent monitor reading is no longer valid for
| |
| * The Emergency Director should declare the Unusual Event promptly upon
| |
| * The Emergency Director should declare the event promptly upon determining classification purposes." This allows for more timely classification since all the basis determining that 60 minutes has been exceeded, or will likely be that the applicable time has been exceeded, or will likely be exceeded. information pertaining to Note bullet 3 will be contained in the IC and therefor exceeded. readily available on the 1 x17 procedure matrix used by the SM.
| |
| * If an ongoing release is detected and the release start time is unknown, assume
| |
| * If an ongoing release is detected and the release start time is unknown, that the release duration has exceeded 60 minutes.
| |
| assume that the release duration has exceeded 60 minutes.
| |
| * Classification based on effluent monitor readings assumes that a release path
| |
| * If the effluent flow past an effluent monitor is known to have stopped, to the environment is established. If the effluent flow past an effluent monitor is indicating that the release path is isolated, the effluent monitor reading is known to have stopped due to actions to isolate the release path, then the no longer valid for classification purposes. effluent monitor reading is no longer valid for classification purposes.
| |
| : 1. Reading on ANY effluent radiation monitor greater than 2 times the (site- 1. Reading on ANY of the following effluent monitors > 2 times alarm specific effluent release controlling document) limits for 60 minutes or setpoint established by a current radioactive release discharge permit longer: for > 60 minutes.
| |
| (site-specific monitor list and threshold values corresponding to 2 times
| |
| * Radwaste Effluent Monitor 1/2-1799-01 the controlling document limits)
| |
| OR
| |
| * Discharge Permit specified monitor
| |
| : 2. Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for OR 60 minutes or longer. 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS
| |
| - Total Noble Gas Release Rate).
| |
| : 3. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times (site-specific effluent release OR controlling document limits) for 60 minutes or longer. 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
| |
| Page 4 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification AG2 RG2 Initiating Condition -- GENERAL EMERGENCY Initiating Condition: F-- No Change E Difference 1 Deviation Spent fuel pool level cannot be restored to at least (site-specific Level 3 description)
| |
| Spent fuel pool level cannot be restored to at least (site-specific Level 3 for 60 minutes or longer.
| |
| description) for 60 minutes or longer. 1) EAL will be used in accordance with the discussion in Section 1.4, NRC Order Operating Mode Applicability: EA-12-051, it is recommended that this EAL be implemented when the enhanced Operating Mode Applicability: All spent fuel pool level instrumentation is available for use. The completion of the 1, 2, 3,4, 5, D enhanced SFP level indicators and need for the inclusion of this EAL is being Example Emergency Action Levels: Emergency Action Level (EAL): tracked in accordance with Exelon Generation Company, LLC's Initial Status Report to March 12, 2012 Commission Order Modifying Licenses with Regard for NOTES: The Emergency Director should declare the General Emergency Note: The Emergency Director should declare the General Emergency promptly Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated promptly upon determining that 60 minutes has been exceeded, upon determining that the applicable time has been exceeded, or will likely October 25,2012.
| |
| or will likely be exceeded be exceeded.
| |
| : 1. Spent fuel pool level cannot be restored to at least (site-specific Level 3 Spent fuel pool level cannot be restored to at least (site-specific Level 3 value) for description) for 60 minutes or longer. 60 minutes or longer.
| |
| Page 5 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition - SITE AREA EMERGENCY Initiating Condition: 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 Order Operating Mode Applicability: All Operating Mode Applicability: EA-12-051, it is recommended that this EAL be implemented when the enhanced 1, 2, 3, 4, 5, D spent fuel pool level instrumentation is available for use. The completion of the Example Emergency Action Levels: enhanced SFP level indicators and need for the inclusion of this EAL is being Emergency Action Level (EAL): tracked in accordance with Exelon Generation Company, LLC's Initial Status
| |
| : 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). 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 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification AA2 RA2 Initiating Condition - ALERT Significant lowering of water level above, or damage to, irradiated fuel.
| |
| Initiating Condition: D No Change f Difference 1 Deviation Significant lowering of water level above, or damage to, irradiated fuel. 1) Listed site-specific monitors and Threshold values to ensure timely classification.
| |
| Operating Mode Applicability: All Operating Mode Applicability:
| |
| : 2) EAL #3 will be used in accordance with the discussion in Section 1.4, NRC 1, 2, 3,4, 5, D Order EA-1 2-051, it is recommended that this EAL be implemented when the Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): 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
| |
| : 1. Uncovery of irradiated fuel in the REFUELING PATHWAY. 1. Uncovery of irradiated fuel in the REFUELING PATHWAY. tracked in accordance with Exelon Generation Company, LLC's Initial Status OR Report to March 12, 2012 Commission Order Modifying Licenses with Regard for
| |
| : 2. Damage to irradiated fuel resulting in a release of radioactivity from the Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) dated fuel as indicated by ANY of the following radiation monitors: 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel October 25,2012.
| |
| as indicated by ANY Table RI Radiation Monitor reading >1000 mRem/hr.
| |
| (site-specific listing of radiation monitors, and the associated readings, OR setpoints and/or alarms)
| |
| : 3. Lowering of spent fuel pool level to (site-specific Level 2 value).
| |
| : 3. Lowering of spent fuel pool level to (site-specific Level 2 value).
| |
| 4.
| |
| Table RI Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Monitor
| |
| * 1(2) 1705-168 Fuel Pool Rad Monitor Page 7 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL [ Justification AU2 RU2 Initiating Condition: UNUSUAL EVENT Initiating Condition: E No Change E Difference 1
| |
| -- Deviation UNPLANNED loss of water level above irradiated fuel. UNPLANNED loss of water level above irradiated fuel 1) Listed site specific level indication and monitors to ensure timely Operating Mode Applicability: All Operating Mode Applicability: classification.
| |
| 1, 2, 3, 4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| : 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by indicated by ANY of the following: ANY of the following:
| |
| * Refueling Cavity water level < 282 in. (Upper Wide range (site-specific level indications). simulated signal).
| |
| OR AND
| |
| * Spent Fuel Pool water level <19 ft. above the fuel (<-4 ft.
| |
| indicated level).
| |
| OR
| |
| : b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
| |
| * Indication or report of a drop in water level in the REFUELING PATHWAY.
| |
| AND (site-specific list of area radiation monitors)
| |
| : b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table RI.
| |
| Table R1 Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Mon
| |
| * 1(2) 1705-16B Fuel Pool Rad Mon Page 8 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification RA3 A3 E No Change 1 Difference 1 Deviation Initiating Condition:
| |
| Initiating Condition - ALERT Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or 1) Listed site specific plant rooms and areas with identified mode Radiation levels that impede access to equipment necessary for normal shutdown. applicability to ensure timely classification.
| |
| plant operations, cooldown or shutdown.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability: All 1, 2, 3, 4, 5,D Emergency Action Level (EAL):
| |
| Example Emergency Action Levels: (1 or 2)
| |
| Note:
| |
| Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no If the equipment in the room or area listed in Table R3 was already inoperable, or out of emergency classification is warranted service, before the event occurred, then no emergency classification is warranted.
| |
| : 1. Dose rate greater than 15 mR/hr in ANY of the following areas:
| |
| * Control Room 1. Dose rate> 15 mR/hr in ANY of the following Table R2 areas:
| |
| * Central Alarm Station Table R2 Areas Requiring Continuous Occupancy
| |
| * (other site-specific areas/rooms)
| |
| * Main Control Room (Unit 1 ARM Station #22)
| |
| : 2. An UNPLANNED event results in radiation levels that prevent
| |
| * Central Alarm Station - (by survey) or significantly impede access to any of the following plant rooms or areas: OR (site-specific list of plant rooms or areas with entry-related 2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to mode applicability identified) ANY of the following Table R3 plant rooms or areas:
| |
| Table R3 Areas with Entry Related Mode Applicability Entry Related Area Unit Mode Applicability Reactor Building
| |
| * First Floor North Wall 1
| |
| * Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Comer 2 Page 9 of 66
| |
| | |
| Justification Rev 66 Proposed EAL NEI 99-01 Rev NEI 99-01 Proposed EAL Justification SU3 RU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: H No Change E Difference I: Deviation Reactor coolant activity greater than Technical Specification allowable limits. Reactor coolant activity greater than Technical Specification allowable limits. 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| : 2) Listed this system category EAL in the radiological category EAL section to Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 maintain consistency with current and previous revisions of Exelon EALs. This will ensure a timely classification since the threshold values are more aligned with the Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| radiological category vice system category.
| |
| : 1. (Site-specific radiation monitor) reading greater than (site-specific value). 1. Offgas system radiation monitor HI-HI alarm.
| |
| OR OR
| |
| : 2. Sample analysis indicates that a reactor coolant activity value is greater than 2. Specific coolant activity > 4.0 uCi/gm Dose equivalent 1-131. -
| |
| an allowable limit specified in Technical Specifications.
| |
| Page 10 of 66
| |
| | |
| NEI 99-01 rev 6 Fission Product Barrier Matrix Fission____________
| |
| Prdc Mari _________________ Hot________
| |
| FGI Loss of any two barriers AND Loss or Potential Loss of third barrier. 1,2,3 FS1 Loss or Potential Loss of ANY two barriers. 1,2,3 FAI ANY Loss or ANY Potential Loss of either Fuel Clad or RCS 1,2,3 FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A. UNPLANNED rapid drop in A. Primary containment pressure
| |
| : 1. RCS Activity primary containment pressure greater than (site-specific value)
| |
| Primary A. (Site specific indications that A. Primary containment pressure follouin rime Containment reactor coolant activi is None greater than (site-specific None B. (site-specific explosive mixture)
| |
| Pressure / greater than 300 uCi/gm value) due to RCS leakage. OR exists inside primary containment Primary dose equivalent 1-131)exssiidprmycotnet Containment B. Primary containment pressure OR Conditions response not consistent with LOCA conditions. C. HCTL exceeded.
| |
| A. RPV water level cannot be A. RPV water level cannot be A. Primary containment flooding restored and maintained above restored and maintained above
| |
| : 2. RPV Water required. (site-specific RPV water level (site-specific RPV water level None None Level corresponding to top of active corresponding to the top of required.
| |
| active fuel) or cannot be fuel) or cannot be determined, determined.
| |
| A. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal OR A. UNISOLABLE break in any of the A. UNISOLABLE primary system containment following: ( site-specific systems leakage thath results EIHRo in exceeding olwn:venting perprimary B. Intentional EOPs 3.RCS Leak with potential for high-energy line EITHER of the following:
| |
| Rate/ Primary None None breas) 1. Max Normal Operating OR None Containment OR Temperature C. UNISOLABLE primary system Isolation Failure B. Emergency RPV OR leakage that results in exceeding Depressurization 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 A. Primary Containment Radiation A. Primary Containment Radiation Containment Radiation Monitor reading None Monitor reading greater than None None Monitor reading greater than (site-Raiation greater than (site-specific (site-specific value).
| |
| Radiation specific value).
| |
| value).
| |
| : 5. Emergency A. Any Condition in the opinion A.Any Condition in the opinion of A. ANY Condition in the opinion of A. Any Condition in the opinion of the A.Any Condition in the opinion of the A. Any Condition in the opinion of the Director of the Emergency Director the Emergency Director that the Emergency Director that Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates that indicates Loss of the indicates Potential Loss of the indicates Loss of the RCS Potential Loss of the RCS Barrier. Loss of the Containment Barrier. Potential Loss of the Containment Judgment Fuel Clad Barrier. Fuel Clad Barrier. Barrier. P I Barrier.
| |
| Page II of 66
| |
| | |
| PrnrnnserlPinnirin Prorlirot flurries Matrin Fl Loss of any two barriers AND Loss or Potential Loss of third barrier. PSi Loss or Potential Loss of ANY two barriers. EFAl ANY Loss or ANY Potential Loss of either Fuel Clad or RCS i FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity > 300 uCi/gm Dose None Nuns
| |
| : 1. RCS Activity Equivalent 1-131. NnNoeNone None None
| |
| : 2. RPV water level cannot be restored and 1. RPV water level cannot be restored and 21 RPV Water 1. Plant conditions indicate Primary maintained > -142 inches (TAF) maintained >-142 inches (TAF) Plant conditions indicate Primary Containment 2Rver Containment flooding is required. OR OR None None flooding is required.
| |
| Level 3. RPV water level cannot be determined. 2. RPV water level cannot be determined.
| |
| : 3. Primary Containment pressure > 56 pslg.
| |
| OR
| |
| : 1. UNPLANNED rapid drop in Drywell 4. a. Drywellor torus hydrogen concentration
| |
| : 1. Drywallpressure '2.5 psig. pressure followingprimary containment > 6%.
| |
| : 3. Primary AND pressure rise. AND Containment None None 2. Drywellpressure rise is due to RCS None OR b. Drywellor tolrs oxygen concentration Pressure/Conditions leakage 2. Drywellpressure response not consistent _0%.
| |
| with LOCAconditions. OR
| |
| : 5. Heat Capacity Limit(OGA200. Figure M) exceeded.
| |
| : 3. UNISOLABLE primary system leakage that
| |
| : 1. UNISOLABLE MainSteam Line (MSL), results in EITHER of the following:
| |
| HPCI. Feedwater. RWCU or RCIC line a. Secondary Containment area break. temperature n OGA 300 Maximum 4.RCS Leak Rate None None OR Normal operating levels.
| |
| : 2. Emergency RPV Depressurization is OR required. b.Secondary Containment area radiation level > QGA300 Maximum Normeal operating level.
| |
| 5.Primary Drywellradiation monitor reading Drywellradiation monitor reading Drywellradiation monitor reading Containment > 6.65 Ev02 R/hr. None >10OR/thr (>t.00 E+02 R/hr). None None > 1.55 E+03 Rthr Radiation
| |
| : 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 ofthe following:
| |
| : a. Secondary Containment area temperature > QGA300, Maximum Safe operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > OGA 300, Maximum Safe operating levels
| |
| : 7. Emergen A.AnyCondition in the opinion of the A. Any Condition in the opinion of the A.ANYCondition in the opinion of the A.AnyCondition in the opinion of the A.Any Condition in the opinion of the A.Any Condition in the opinion of the 7.Eergecy 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 Barner. of the Containment Barrier.
| |
| Page 12 of66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification FCl FC1 Category: Fuel Clad Barrier Category: Fuel Clad Barrier E No Change -- 1 Difference 1 Deviation RCS Activity RCS Activity 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. (Site specific indications that reactor coolant activity is greater than 300 uCi/gm Coolant activity > 300 uCi/gm Dose Equivalent 1-131.
| |
| dose equivalent 1-131)
| |
| Page 13 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification FC2 FC2 I-- No Change 1 Difference 1 Deviation Category: Fuel Clad Barrier Category: Fuel Clad Barrier RPV Water Level RCS Activity 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary containment flooding required. 1. Plant conditions indicate Primary Containment flooding is required.
| |
| Potential Loss Potential Loss A. RPV water level cannot be restored and maintained above (site-specific RPV 2. RPV water level cannot be restored and maintained > -142 inches (TAF) water level corresponding to top of active fuel) or cannot be determined.
| |
| OR
| |
| : 3. RPV water level cannot be determined.
| |
| Page 14 of66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification FC4 FC5 Category: Fuel Clad Barrier Category: Fuel Clad Barrier No Change 1 Difference 1 Deviation Primary Containment Radiation Primary Containment Radiation 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywall radiation monitor reading > 6.65 E+02 R/hr.
| |
| value).
| |
| Page 15 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification FC6 FC7 Category: Fuel Clad Barrier Category: Fuel Clad Barrier 1-- No Change D Difference 1 Deviation Emergency Director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. Any Condition in the opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier. the Fuel Clad Barrier.
| |
| Potential Loss Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier. Loss of the Fuel Clad Barrier.
| |
| Page 16 of 66
| |
| | |
| NEI 99-01 Rev 6 j Proposed EAL Justification RCI RC3 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change D Difference 1: Deviation Primary Containment Pressure Primary Containment Pressure/Conditions 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary containment pressure greater than (site-specific value) due to RCS 1. Drywell pressure >2.5 psig.
| |
| leakage. AND
| |
| : 2. Drywell pressure rise is due to RCS leakage.
| |
| Page 17 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification RC2 RC2 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change L Difference L 1 Deviation RPV Water Level RPV Water Level 1) Listed site-specific threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. RPV water level cannot be restored and maintained above (site-specific RPV 1. RPV water level cannot be restored and maintained > -142 inches (TAF).
| |
| level corresponding to the top of active fuel) or cannot be determined. OR
| |
| : 2. RPV water level cannot be determined.
| |
| Page 18 of66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification RC3 RC4 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier W No Change 1 Difference 1 Deviation RCS Leak Rate RCS Leak Rate 1) Listed site-specific systems and threshold values to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. UNISOLABLE break in any of the following: ( site-specific systems with potential 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater,, RWCU or RCIC for high-energy line breas) line break.
| |
| OR OR
| |
| : 2. Emergency RPV Depressurization is required.
| |
| B. Emergency RPV Depressurization Potential Loss Potential Loss A. UNISOLABLE primary system leakage that results in exceeding EITHER of the following: 3. UNISOLABLE primary system leakage that results in EITHER of the following:
| |
| : a. Secondary Containment area temperature > QGA 300 Maximum Normal
| |
| : 1. Max Normal Operating Temperature. operating levels.
| |
| OR OR
| |
| : b. Secondary Containment area radiation level > QGA 300 Maximum Normal
| |
| : 2. Max Normal Operating Area Radiation Level.
| |
| operating levels.
| |
| Page 19 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification RC4 RC5 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change 1 Difference 1: Deviation Primary Containment Radiation Primary Containment Radiation 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 100 R/hr (>1.00E+02 R/hr).
| |
| value).
| |
| Page 20 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification RC6 RC7 Category: Reactor Coolant System Barrier Category: Reactor Coolant System Barrier E No Change D Difference 1 Deviation Emergency Director Judgment Emergency director Judgment Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. ANY Condition in the opinion of the Emergency Director that indicates Loss of the RCS Barrier. the RCS Barrier.
| |
| Potential Loss Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier. 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the RCS Barrier.
| |
| Page 21 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification CT1 CT3 Category: Containment Barrier Category: Containment Barrier D No Change E Difference I: Deviation Primary Containment Conditions Primary Containment Pressure/Conditions 1) Listed site-specific threshold values to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| : 2) The words "and rising" were added to account for the momentary spike in Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 pressure where pressure is now lowering, the risk of a potential loss of Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold: containment is no longer present, this wording is also consistent with present EAL wording.
| |
| Loss Loss C. UNPLANNED rapid drop in primary containment pressure following primary 1. UNPLANNED rapid drop in Drywell pressure following Drywall pressure rise.
| |
| containment pressure rise OR OR 2. Drywell pressure response not consistent with LOCA conditions.
| |
| B. Primary containment pressure response not consistent with LOCA conditions. Potential Loss Potentiat Loss
| |
| : 3. Drywell pressure > 56 psig and rising.
| |
| D. Primary containment pressure greater than (site-specific value)
| |
| OR OR 4. a. Drywell or torus Hydrogen concentration > 6%.
| |
| E. (site-specific explosive mixture) exists inside primary containment AND OR b. Drywell or torus Oxygen concentration > 5%.
| |
| OR
| |
| : 3. HCTL exceeded. 5. Heat Capacity Limit (QGA 200, Fig. M) exceeded.
| |
| Page 22 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CT2 CT2 Category: Containment Barrier Category: Containment Barrier Ejj No Change jjj Difference 1-1 Deviation RPV Water Level RPV Water Level Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Potential Loss Potential Loss A. Primary containment flooding required.
| |
| Plant conditions indicate Primary Containment flooding is required.
| |
| Page 23 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification CT3 CT6 Category: Containment Barrier Category: Containment Barrier 1fl No Change E Difference r-- Deviation Primary Containment Isolation Failure Primary Containment Isolation Failure 1) Listed site-specific threshold values to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. UNISOLABLE direct downstream pathway to the environment exists after 1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal primary containment isolation signal.
| |
| OR OR
| |
| : 2. Intentional Primary Containment venting/purging per EOPs or SAMGs due to B. Intentional primary containment venting per EOPs accident conditions.
| |
| OR OR C. UNISOLABLE primary system leakage that results in exceeding EITHER of the 3. UNISOLABLE primary system leakage that results in EITHER of the following:
| |
| following: a. Secondary Containment area temperature > QGA 300 Maximum Safe operating levels.
| |
| : 1. Max Safe Operating Temperature. OR OR b. Secondary Containment area radiation level > QGA 300 Maximum Safe
| |
| : 2. Max Safe Operating Area Radiation Level. operating levels.
| |
| Page 24 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CT4 CT5 Category: Containment Barrier Category: Containment Barrier No Change 1 Difference 1 Deviation Primary Containment Radiation Primary Containment Radiation 1) Listed site-specific monitor and threshold value to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Potential Loss Potential Loss A. Primary Containment Radiation Monitor reading greater than (site-specific Drywell radiation monitor reading > 1.55 E+03 R/h.
| |
| value).
| |
| Page 25 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CT6 CT7 Category: Containment Barrier No Change Difference Deviation Category: Containment Barrier Emergency director Judgment Emergency Director Judgment Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Fission Product Barrier Threshold: Fission Product Barrier (FPB) Threshold:
| |
| Loss Loss A. Any Condition in the opinion of the Emergency Director that indicates Loss of 1. Any Condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier. the Containment Barrier.
| |
| Potential Loss Potential Loss A. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier. 2. Any Condition in the opinion of the Emergency Director that indicates Potential Loss of the Containment Barrier.
| |
| Page 26 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SG1 MG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition:
| |
| I-* No Change EIDifference -- 1 Deviation Prolonged loss of all offsite and all onsite AC power to emergency buses. Prolonged loss of all offsite and all onsite AC power to emergency buses.
| |
| : 1) Listed site specific equipment, site specific time based on station blackout Operating Mode Applicability: Operating Mode Applicability: coping analysis, and site specific indication to ensure timely classification.
| |
| 1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the General Emergency promptly upon determining that (site-specific hours) has been exceeded, or will Note:
| |
| likely be exceeded.
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| : 1. a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).
| |
| 1 Loss of ALL offsite AC power to unit ECCS busses.
| |
| AND AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel
| |
| : b. EITHER of the following: generators to supply power to unit ECCS busses.
| |
| AND
| |
| * Restoration of at least one emergency bus in less than (site-specific hours) is not likely. 3. EITHER of the following:
| |
| : a. Restoration of at least one unit ECCS bus in < I hour is not likely.
| |
| * (Site-specific indication of an inability to adequately OR remove heat from the core)
| |
| : b. RPV water level cannot be restored and maintained > -166 Inches.
| |
| Page 27 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification SS1 MS1 Initiating Condition: SITE AREA EMERGENCY Initiating Condition:
| |
| Loss of all offsite and onsite AC power to emergency buses for 15 minutes or M No Change D Difference FIDeviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.
| |
| longer.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| 1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Site Area Emergency Note:
| |
| promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) exceeded.
| |
| for 15 minutes or longer.
| |
| : 1. Loss of ALL offsite AC Power to unit ECCS buses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Page 28 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification SA1I MA1I Initiating Condition: ALERT Initiating Condition:
| |
| -<I No Change [ ]Difference F Deviation Loss of all but one AC power source to emergency buses for 15 minutes or longer. Loss of all but one AC power source to emergency buses for 15 minutes or longer.
| |
| : 1) Listed site specific equipment to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:
| |
| : 1. a. AC power capability to (site-specific emergency buses) is reduced to a The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be single power source for 15 minutes or longer. exceeded.
| |
| AND
| |
| : 1. AC power capability to unit ECCS buses reduced to only one of the following
| |
| : b. Any additional single power source failure will result in loss of all AC power sources for > 15 minutes.
| |
| power to SAFETY SYSTEMS.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit Auxiliary Transformer TR-11 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers AND
| |
| : 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.
| |
| Page 29 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification MUl Initiating Condition: UNUSUAL EVENT Initiating Condition:
| |
| FI No Change EIDifference F- Deviation Loss of all offsite AC power capability to emergency buses for 15 minutes or Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. longer. 1) Listed site specific equipment to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 exceeded.
| |
| minutes or longer Loss of ALL offsite AC power capability to unit ECCS buses for > 15 minutes.
| |
| Page 30 of 66
| |
| | |
| NEI 99-01 Rev 6 J Proposed EAL Justification Initiating Condition: GENERAL EMERGENCY Loss of all AC and Vital DC power sources for 15 minutes or longer.
| |
| SG8 Initiating Condition:
| |
| MG2 D No Change M Difference I Deviation Loss of all AC and Vital DC power sources for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.
| |
| Operating Mode Applicability: 2) in EAL #3 removed the word "indicated" this will allow for an indication problem Operating Mode Applicability: to not cause confusion on the need to declare.
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded.
| |
| : 1. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be buses) for 15 minutes or longer. exceeded.
| |
| AND
| |
| : 1. Loss of ALL offsite AC power to unit ECCS busses.
| |
| Indicated voltage is less than (site-specific bus voltage value) on ALL AND (site-specific vital DC buses) for 15 minutes or longer.
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.
| |
| AND
| |
| : 3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.
| |
| AND
| |
| : 4. ALL AC and Vital DC power sources have been lost for > 15 minutes.
| |
| Page 31 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification ss8 MS2 Initiating Condition: SITE AREA EMERGENCY Loss of all Vital DC power for 15 minutes or longer.
| |
| Initiating Condition: E] No Change H- Difference 1- Deviation Loss of all Vital DC power for 15 minutes or longer.
| |
| : 1) Listed site specific equipment and value to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability: 2) Removed the word "indicated" this will allow for an indication problem to not Power Operation, Startup, Hot Standby, Hot Shutdown cause confusion on the need to declare.
| |
| 1,2,3 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:
| |
| Indicated voltage is less than (site-specific bus voltage value) on ALL Vital DC
| |
| * The Emergency Director should declare the event promptly upon buses for 15 minutes or longer. determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| Voltage is < 105 VDC on 125 VDC battery buses #1 and #2 for >15 minutes..
| |
| Page 32 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SS5 MS3 Initiating Condition: SITE AREA EMERGENCY Initiating Condition: D No Change FIDifference F Deviation Inability to shutdown the reactor causing a challenge to (core cooling [PWR] / RPV water level [BWR]) or RCS heat removal. Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. 1) Listed site specific indications to ensure timely classification.
| |
| Operating Mode Applicability: 2) Mode 2 included in operating mode applicability as per developer notes.
| |
| Operating Mode Applicability:
| |
| Power Operation 1,2 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| : 1. a. An automatic (trip [PWRI / scram [BWR]) did not shutdown the reactor.
| |
| : 1. Automatic scram did not shutdown the reactor as indicated by Reactor Power AND > 5%.
| |
| AND
| |
| : b. All manual actions to shutdown the reactor have been unsuccessful. 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
| |
| AND AND
| |
| : c. EITHER of the following conditions exist: 3. EITHER of the following conditions exist:
| |
| : 1. (Site-specific indication of an inability to adequately " RPV water level cannot be restored and maintained > -166 inches.
| |
| remove heat from the core) OR
| |
| " Heat Capacity Limit (CGA 200, Figure M) exceeded.
| |
| OR
| |
| : 2. (Site-specific indication of an inability to adequately remove heat from the RCS)
| |
| Page 33 of 66
| |
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| NEI 99-01 Rev 6 Proposed EAL Justification SA5 MA3 Initiating Condition: ALERT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and Initiating Condition: F No Change M Difference [-] Deviation Automatic or manual scram fails to shutdown the reactor, and subsequent manual 1) Listed site specific indications to ensure timely classification.
| |
| subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. actions taken at the reactor control consoles are not successful in shutting down the reactor. 2) Mode 2 included in operating mode applicability as per developer notes.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation 1,2 Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually Note:
| |
| driving in control rods or implementation of boron injection strategies.
| |
| A manual action is any operator action, or set of actions, which causes the
| |
| : 1. a. An automatic (trip [PWR] / scram [BWR]) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor.
| |
| manually driving in control rods or implementation of boron injection strategies.
| |
| AND
| |
| : 1. Automatic or manual scram did not shutdown the reactor as indicated by
| |
| : b. Manual action taken at the reactor control consoles are not Reactor Power > 5%.
| |
| successful in shutting down the reactor.
| |
| AND
| |
| : 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.
| |
| Page 34 of 66
| |
| | |
| Justification Proposed EAL NEI Rev 66 99-01 Rev NEt 99-01 Proposed EAL Justification SU5 MU3 Initiating Condition: UNUSUAL EVENT Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor.
| |
| Initiating Condition: F No Change W Difference FI Deviation Automatic or manual scram fails to shutdown the reactor.
| |
| : 1) Listed site specific indications to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| : 2) Mode 2 included in operating mode applicability as per developer notes.
| |
| Power Operation 1,2 Example Emergency Action Levels: (1 or 2)
| |
| Emergency Action Level (EAL):
| |
| Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note:
| |
| A manual action is any operator action, or set of actions, which causes the
| |
| : 1. a. An automatic (trip [PWRJ / scram [BWR]) did not shutdown the control rods to be rapidly inserted into the core, and does not include reactor. manually driving in control rods or implementation of boron injection strategies.
| |
| AND
| |
| : a. Automatic scram did not shutdown the reactor as indicated by Reactor
| |
| : b. A subsequent manual action taken at the reactor control consoles is Power > 5%.
| |
| successful in shutting down the reactor.
| |
| AND
| |
| : b. Subsequent manual / ARI action taken at the Reactor Console is
| |
| : 2. a. A manual scram ([PWR] / scram [BWR]) did not shutdown the successful in shutting down the reactor.
| |
| reactor.
| |
| OR AND 2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power > 5%.
| |
| : b. EITHER of the following: AND
| |
| : b. EITHER of the following:
| |
| : 1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. 1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor.
| |
| OR OR
| |
| : 2. Subsequent automatic scram / ARI is successful in shutting down
| |
| : 2. A subsequent automatic (trip [PWR] / scram [BWRJ) is successful the reactor.
| |
| in shutting down the reactor.
| |
| Page 35 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SA2 MA4 Initiating Condition: ALERT Initiating Condition: F- No Change DIDifference F Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Note: The Emergency Director should declare the Alert promptly upon Emergency Action Level (EAL):
| |
| determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded.
| |
| The Emergency Director should declare the event promptly upon
| |
| : 1. a. An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.
| |
| [see table below]
| |
| : 1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter
| |
| [ BWR parameter list] [PWR parameter list] from within the Control Room for >15 minutes.
| |
| Reactor Power Reactor Power Table M1 Control Room Parameters RPV Level RCS Level
| |
| " Reactor Power
| |
| " RPV Water Level RPV Pressure RCS Pressure " RPV Pressure
| |
| " Drywell Pressure Primary Containment Pressure In Core/Core Exit Temperature " Torus Level
| |
| " Torus Temperature Levels in at least (site specific Suppression Pool Level number) steam generators AND Suppression Pool Temperature Steam Generator Auxiliary or 2. ANY Table M2 transient in progress.
| |
| Emergency Feed Water Flow AND Table M2 Significant Transients
| |
| : b. Any of the following transient events in progress. " Turbine Trip
| |
| " Automatic or Manual runback greater than 25% thermal reactor power " Reactor Scram
| |
| " Electrical load rejection greater than 25% full electrical load " ECCS Activation
| |
| " Reactor Scram [BWR] / trip [PWR]
| |
| " ECCS (SI) actuation " Recirc. Runback > 25% Reactor Power Change
| |
| " Thermal power oscillations greater than 10% [BWR] " Thermal Power oscillations > 10% Reactor Power Change Page 36 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SU2 MU4 Initiating Condition: UNUSUAL EVENT UNPLANNED loss of Control Room indications for 15 minutes or longer.
| |
| Initiating Condition: F- No Change IDifference F Deviation UNPLANNED loss of Control Room indications for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 Example Emergency Action Levels:
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):
| |
| determining that 15 minutes has been exceeded, or will likely be Note:
| |
| exceeded.
| |
| The Emergency Director should declare the event promptly upon An UNPLANNED event results in the inability to monitor one or more of the determining that the applicable time has been exceeded, or will likely be following parameters from within the Control Room for 15 minutes or longer. exceeded.
| |
| [see table below] UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for > 15 minutes.
| |
| [ BWR parameter list] [ PWR parameter list]
| |
| Table MI Control Room Parameters Reactor Power Reactor Power
| |
| " Reactor Power RPV Level RCS Level " RPV Water Level
| |
| " RPV Pressure
| |
| " Drywell Pressure RPV Pressure RCS Pressure
| |
| " Torus Level
| |
| " Torus Temperature Primary Containment Pressure InCore(Core Exit Temperature Levels in at least (site specific Suppression Pool Level number) steam generators Suppression nPool Temperature PEmergency Steam Generator Auxiliary Feed Water or Flow Page 37 of66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification SA9 MA5 Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM needed for the current operating Initiating Condition: ALERT Hazardous event affecting a SAFETY SYSTEM required for the current operating FD1No Change
| |
| * Difference F Deviation
| |
| : 1) No additional site specific hazard noted mode. mode.
| |
| : 2) Changed the word "needed" to "required" in the IC and "required by Technical Operating Mode Applicability: Operating Mode Applicability: Specification" in the EAL to be consistent with terminology used by operators and minimize confusion.
| |
| 1,2,3 Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Level (EAL): 3) Added additional note to easily direct the operator to lesser ICs Note:
| |
| Example Emergency Action Levels:
| |
| If it is determined that the conditions of MA5 are not met then assess the
| |
| : 1. a. The occurrence of ANY of the following hazardous events: event via HU3, HU4, or HU6.
| |
| * Seismic event (earthquake)
| |
| The occurrence of ANY of the following hazardous events:
| |
| * Internal or extemal flooding event
| |
| * Seismic event (earthquake)
| |
| * High winds or tornado strike
| |
| * FIRE " Internal or external flooding event
| |
| * EXPLOSION
| |
| * High winds or tornado strike
| |
| * (site-specific hazards)
| |
| * FIRE
| |
| * Other events with similar hazard characteristics as determined by the Shift Manager
| |
| * EXPLOSION
| |
| * Other events with similar hazard characteristics as determined AND by the Shift Manager AND
| |
| : b. EITHER of the following:
| |
| : 2. EITHER of the following:
| |
| : 1. Event damage has caused indications of degraded performance a. Event damage has caused indications of degraded in at least one train of a SAFETY SYSTEM needed for the performance in at least one train of a SAFETY SYSTEM current operating mode. required by Technical Specifications for the current operating mode.
| |
| OR OR
| |
| : b. The event has caused VISIBLE DAMAGE to a SAFETY
| |
| : 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM SYSTEM component or structure required by Technical component or structure needed for the current operating mode.
| |
| Specifications for the current operating mode.
| |
| Page 38 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification SU4 MU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change FIDifference FIDeviation RCS leakage for 15 minutes or longer.
| |
| RCS leakage for 15 minutes or longer. 1) Listed site specific values to ensure timely classification.
| |
| Operating Mode Applicability: 2) Changed wording from containment to Drywall for clarity to better define the Operating Mode Applicability: primary containment structure.
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1,2,3 3) In EAL #1 and 2 added "into the Drywell" to differentiate between EAL #1/2 and
| |
| #3. Without this wording would have been in EAL #1 or #2 concurrent with #3. With Example Emergency Action Levels: (1 or 2 or 3) the added wording each EAL can be called separately.
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Note:
| |
| determining that 15 minutes has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be
| |
| : 1. RCS unidentified or pressure boundary leakage greater than (site-specific exceeded.
| |
| value) for 15 minutes or longer.
| |
| : 1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for
| |
| : 2. RCS identified leakage greater than (site-specific value) for 15 minutes or > 15 minutes.
| |
| longer OR
| |
| : 2. RCS identified leakage in the Drywell >25 gpm for > 15 minutes.
| |
| : 3. Leakage from the RCS to a location outside containment greater than 25 OR gpm for 15 minutes or longer
| |
| : 3. Leakage from the RCS to a location outside the Drywell >25 gpm for
| |
| > 15 minutes.
| |
| Page 39 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification SU6 Initiating Condition: UNUSUAL EVENT Loss of all onsite or offsite communications capabilities Initiating Condition:
| |
| MU7 W No Change E Difference F Deviation Loss of all onsite or offsite communication capabilities. 1) Listed site specific communication methods to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Power Operation, Startup, Hot Standby, Hot Shutdown 1.2,3 Example Emergency Action Levels: (1 or 2 or 3)
| |
| Emergency Action Level (EAL):
| |
| : 1. Loss of ALL of the following onsite communication methods:
| |
| (site-specific list of communications method 1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine operations.
| |
| : 2. Loss of ALL of the following ORO communications s) methods: OR
| |
| : 2. Loss of ALL Table M3 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.
| |
| OR
| |
| : 3. Loss of ALL of the following NRC communications methods:
| |
| : 3. Loss of ALL Table M3 NRC communication capability affecting the ability to perform NRC notifications.
| |
| (site-specific list of communications methods)
| |
| Table M3 Communications Capabil'y -
| |
| System Onsite Offsite NRC Plant Radio X Plant Page X All telephone Lines (Commercial and X X X microwave)
| |
| ENS X X HPN X X Satellite Phones X X Page 40 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification CA2 CAl Initiating Condition: ALERT Initiating Condition: E No Change 1 Difference I: Deviation Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or Loss of all offsite and onsite AC power to emergency buses for 15 minutes or longer. 1) Listed site specific equipment to ensure timely classification.
| |
| longer.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| 4,5, D Cold Shutdown, Refueling, Defueled Emergency Action Level (EAL):
| |
| Example Emergency Action Levels:
| |
| Note:
| |
| Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes time has been exceeded, or will likely be The Emergency Director should declare the event promptly upon determining exceeded. that the applicable time has been exceeded, or will likely be exceeded.
| |
| Loss of ALL offsite and ALL onsite AC Power to (site-specific emergency buses) for 15 minutes or longer. 1. Loss of ALL offsite AC power to unit ECCS buses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Page 41 of 66
| |
| | |
| Justification Proposed EAL NEI 99-01 Rev NEI Rev 66 Proposed EAL Justification CU2 Cut Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change 1:1 Difference 1 Deviation Loss of all but one AC power source to emergency buses for 15 minutes or longer. Loss of all but one AC power source to emergency buses for 15 minutes or 1) Listed site specific equipment to ensure timely classification.
| |
| longer.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| Cold Shutdown, Refueling, Defueled 4,5, D Example Emergency Action Levels:
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Emergency Action Level (EAL):
| |
| determining that 15 minutes time has been exceeded, or will likely be exceeded. Note:
| |
| : 1. a. AC power capability to (site-specific emergency buses) is reduced to a The Emergency Director should declare the event promptly upon single power source for 15 minutes or longer. determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| AND
| |
| : 1. AC power capability to unit ECCS buses reduced to only one of the
| |
| : b. Any additional single power source failure will result in loss of all AC following power sources for > 15 minutes.
| |
| power to SAFETY SYSTEMS.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit auxiliary transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers AND
| |
| : 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.
| |
| Page 42 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification Initiating Condition - ALERT CA6 Initiating Condition:
| |
| CA2 D-1 No Change E Difference -- Deviation Hazardous event affecting SAFETY SYSTEM needed for the current operating Hazardous event affecting SAFETY SYSTEM required for the current operating 1) No additional site specific hazards noted mode. mode.
| |
| Operating Mode Applicability: Operating Mode Applicability: 2) Changed the word "needed" to "required" in the IC and "required by Technical Specification" in the EAL to be consistent with terminology used by Cold Shutdown, Refueling 4,5 operators and minimize confusion.
| |
| Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| : 1. a. The occurrence of ANY of the following hazardous events: Note: 3) Added additional note to easily direct the operator to lesser ICs
| |
| " Seismic event (earthquake) If it is determined that the conditions of CA2 are not met then assess the
| |
| " Internal or external flooding event event via HU3, HU4, or HU6.
| |
| * High winds or tornado strike The occurrence of ANY of the following hazardous events:
| |
| " FIRE
| |
| * Seismic event (earthquake) i EXPLOSION
| |
| * Internal or external flooding event
| |
| * (site-specific hazards) " High winds or tornado strike
| |
| " Other events with similar hazard characteristics as
| |
| * FIRE determined by the Shift Manager " EXPLOSION AND
| |
| * Other events with similar hazard characteristics as
| |
| : b. EITHER of the following: determined by the Shift Manager
| |
| : 1. Event damage has caused indications of degraded AND performance in at least one train of a SAFETY SYSTEM 2. EITHER of the following:
| |
| needed for the current operating mode.
| |
| : a. Event damage has caused indications of degraded OR performance in at least one train of a SAFETY SYSTEM
| |
| : 2. The event has caused VISIBLE DAMAGE to a SAFETY required by Technical Specifications for the current SYSTEM component or structure needed for the current operating mode.
| |
| 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.
| |
| Page 43 of 66
| |
| | |
| Justification Proposed EAL NEI 99-01 Rev 6 Proposed EAL Justification CU4 CU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change E Difference [: Deviation Loss of Vital DC power for 15 minutes or longer. Loss of Vital DC power for 15 minutes or longer. 1) Listed site specific voltage and equipment to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| : 2) Removed the word "indicated" this will allow for an indication problem to not Cold Shutdown, Refueling 4,5 cause confusion on the need to declare.
| |
| Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon Note:
| |
| determining that 15 minutes time has been exceeded, or will likely be The Emergency Director should declare the event promptly upon exceeded.
| |
| determining that the applicable time has been exceeded, or will likely be Indicated voltage is less than (site-specific bus voltage value) on required Vital DC exceeded.
| |
| buses for 15 minutes or longer.
| |
| Voltage is < 105 VDC on required 125 VDC battery buses #1 and #2 for > 15 minutes.
| |
| Page 44 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: UNUSUAL EVENT Loss of all onsite or offsite communications capabilities CU5 Initiating Condition:
| |
| CU4 W No Change ED ifference -- 1 Deviation Loss of all onsite or offsite communication capabilities. 1) Listed site specific communications methods to ensure timely classification Operating Mode Applicability: Operating Mode Applicability:
| |
| Cold Shutdown, Refueling, Defuled 4,5, D Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL):
| |
| : 1. Loss of ALL of the following onsite communication methods: 1. Loss of ALL Table C1 Onsite communications capability affecting the (site-specific list of communications method ability to perform routine operations.
| |
| OR
| |
| : 2. Loss of ALL of the following ORO communications s) methods:
| |
| : 2. Loss of ALL Table C1 Offsite communication capability affecting the (site-specific list of communications methods) ability to perform offsite notifications.
| |
| : 3. Loss of ALL of the following NRC communications methods:
| |
| OR (site-specific list of communications methods)
| |
| : 3. Loss of ALL Table C1 NRC communication capability affecting the ability to perform NRC notifications.
| |
| Table Cl Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X All telephone Lines (Commercial and X X X microwave)
| |
| ENS X X HPN X X Satellite Phones X X Page 45 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification CA3 CA5 D NoChange W Difference D eviation Initiating Condition: ALERT Initiating Condition:
| |
| Inability to maintain the plant in cold shutdown. Inability to maintain plant in cold shutdown.
| |
| Operating Mode Applicability: Operating Mode Applicability: 1) Listed site specific Technical Specification cold shutdown temperature limit to ensure timely classification.
| |
| Cold Shutdown, Refueling 4,5
| |
| : 2) Listed site specific pressure reading to enhance timely classification.
| |
| Example Emergency Action Levels: (1 or 2) Emergency Action Levels (EAL):
| |
| Note: The Emergency Director should declare the Alert promptly upon Note: The Emergency Director should declare the event promptly upon 3) Added wording relating the temp and press rise to a loss of decay heat removal determining that the applicable has been exceeded, or will likely be determining that the applicable time has been exceeded, or will likely be capability as per the developer notes for PWR's exceeded. exceeded.
| |
| : 4) Added the following note, taken from the basis section of the EAL, to bring it to
| |
| : 1. UNPLANNED increase in RCS temperature to greater than (site-specific A momentary UNPLANNED excursion above the Technical Specification the attention of the SM/ED when using the "procedure matrix" (11x17 quick Technical Specification cold shutdown temperature limit) for greater than cold shutdown temperature limit when heat removal function is available reference control room document) "Amomentary UNPLANNED excursion above the duration specified in the following table. does not warrant classification. the Technical Specification cold shutdown temperature limit when heat removal 0
| |
| : 2. UNPLANNED RCS pressure increase greater than (site-specific pressure 1. UNPLANNED rise in RCS temperature > 212 F for > Table C2 duration. function is available does not warrant classification."
| |
| reading). (This EAL does not apply during water-solid plant conditions.
| |
| [PWR])
| |
| OR
| |
| : 2. UNPLANNED RCS pressure rise > 10 psig as a result of temperature Table: RCS Heat-up Duration Thresholds rise.
| |
| RCS Status Containment Closure Heat-up Duration Status Table C2 - RCS Heat-up Duration Thresholds Intact (but not RCS Status Containment Closure Heat-up Duration RCS Reduced Not Applicable 60 minutes* Status Inventory (PWR]) Intact Not Applicable 60 minutes*
| |
| Not Intact (or at Established 20 minutes*
| |
| reduced inventory
| |
| [PWR]) Not Established 0 minutes Established 20 minutes*
| |
| * If an RCS heat removal system is in operation within this time frame and Not Intact RCS temperature is being reduced, the EAL is not applicable.
| |
| 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 aonlicable.
| |
| Page 46 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CU3 CU5 Initiating Condition: UNUSUAL EVENT Initiating Condition: H No Change W Difference 1 Deviation UNPLANNED increase in RCS temperature. UNPLANNED rise in RCS temperature. 1) Listed site specific Technical Specification cold shutdown temperature limit Operating Mode Applicability: Operating Mode Applicability: to ensure timely classification.
| |
| Cold Shutdown, Refueling 4,5 2) Changed the word increase to rise in the initiating condition to be consistent Example Emergency Action Levels: (1 or 2) with operations language and training.
| |
| Emergency Action Levels (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly Note: The Emergency Director should declare the event promptly upon 3) Added the following note, taken from the basis section of the EAL, to bring it upon determining that 15 minutes time has been exceeded, or will likely determining that the applicable time has been exceeded, or will likely to the attention of the SM/ED when using the "procedure matrix" (1lx17 quick be exceeded. be exceeded. reference control room document) "A momentary UNPLANNED excursion
| |
| : 1. UNPLANNED increase in RCS temperature to greater than (site-specific above the Technical Specification cold shutdown temperature limit when heat A momentary UNPLANNED excursion above the Technical removal function is available does not warrant classification."
| |
| Technical Specification cold shutdown temperature limit). Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.
| |
| : 2. Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV 0
| |
| [BWRJ) level indication for 15 minutes or longer. 1. UNPLANNED rise in RCS temperature > 212 F.
| |
| OR
| |
| : 2. Loss of the following for > 15 minutes.
| |
| * ALL RCS temperature indications AND
| |
| * ALL RPV level indications Page 47 of66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification CGI CG6 Initiating Condition: GENERAL EMERGENCY Initiating Condition: H No Change E Difference 1-- Deviation Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting fuel clad Loss of reactor vessel / RCS inventory affecting fuel clad integrity with integrity with containment challenged. containment challenged. 1) Listed site specific levels, radiation monitors, and sumps and tanks to ensure Operating Mode Applicability: Operating Mode Applicability: timely classification.
| |
| Cold Shutdown, Refueling 4,5 2) Listed Explosive mixture in the Containment Challenge Table to ensure timely classification.
| |
| Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the General Emergency Note:
| |
| promptly upon determining that 30 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be
| |
| : 1. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) vessel level less than (site- exceeded.
| |
| specific level) for 30 minutes or longer.
| |
| : 1. a. RPV water level < -142 inches (TAF) for > 30 minutes.
| |
| AND AND
| |
| : b. ANY indication from the Containment Challenge Table b. Any Containment Challenge Indication (Table C4)
| |
| : 2. a.. (Reactor vessel/RCS [PWR] or RPV [BWR]) vessel level cannot be OR monitored for 30 minutes or longer. 2. a. RPV water level cannot be monitored for > 30 minutes.
| |
| AND AND
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| * Table C3 indications of a sufficient magnitude to indicate core
| |
| * (Site-specific radiation monitor) reading greater than (site-specific value) uncovery.
| |
| * Erratic source range monitor indication [PWR] OR
| |
| * UNPLANNED increase in (site-specific sump and/or tank levels) of
| |
| * Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.
| |
| sufficient magnitude to indicate core uncovery AND
| |
| * (Other site-specific indications) c. ANY Containment Challenge Indication (Table C4)
| |
| AND Table C3 Indications of RCS Leakage
| |
| : c. ANY indication from the Containment Challenge Table).
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| Table: Containment Challenge Table
| |
| * UNPLANNED Torus level rise*
| |
| * CONTAINMENT CLOSURE not established*
| |
| * (Explosive mixture) exists inside containment
| |
| * UNPLANNED vessel make up rate rise
| |
| * UNPLANNED increase in containment pressure
| |
| * Observation of leakage or inventory loss
| |
| * Secondary containment radiation monitor reading above (site-specific value) [BWR] *Rise in level is attributed to a loss of RPV inventory.
| |
| if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not reouired. Table C4 Containment Challenge Indications
| |
| * Primary Containment Hydrogen Concentration > 6% and Oxygen
| |
| > 5%
| |
| * UNPLANNED rise in containment pressure
| |
| * CONTAINMENT CLOSURE not established*
| |
| * ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.
| |
| if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.
| |
| __
| |
| __ _ __ __ _ _
| |
| __ __ _ _I__ _ _ _ . 'I Page 48 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL I Justification Initiating Condition: SITE AREA EMERGENCY CS1 Initiating Condition:
| |
| CS6 D No Change E Difference 1: Deviation
| |
| : 1) Listed site specific values for level, radiation monitors, and sumps and tanks Loss of (reactor vesselIRCS [PWR1 or RPV [BWR]) inventory affecting core Loss of reactor vessel I RCS inventory affecting core decay heat removal decay heat removal capability. capabilities. to ensure timely classification.
| |
| Operating Mode Applicability: Operating Mode Applicability: 2) EAL #1 Quad Cities Station does not have a low, low, low RPV water level Cold Shutdown, Refueling 4,5 value stated within their Technical Specifications, all ESF functions come from the low, low value of -55.2inches, NEI 99-01 rev 6 developer notes for CSl Example Emergency Action Levels: (1 or 2 or 3) Emergency Action Level (EAL): request the developer to use the low, low, low value for declaration of a SAE Note: The Emergency Director should declare the Site Area Emergency Note: when Containment Closure is not established, as such an RPV water level -65 promptly upon determining that 30 minutes time has been exceeded, or inches was selected as the threshold value since it is substantially above TAF will likely be exceeded.
| |
| The Emergency Director should declare the event promptly upon RPV water level of -142 inches and substantially below the low, low value of determining that the applicable time has been exceeded, or will likely be -55.2inches clearly indicating a worsening of the condition and the need to
| |
| : 1. a. CONTAINMENT CLOSURE not established. exceeded. escalate the emergency from an Alert condition.
| |
| AND
| |
| : b. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (site-
| |
| : 1. With CONTAINMENT CLOSURE not established, RPV water level specific level). < -65 inches
| |
| : 2. a. CONTAINMENT CLOSURE established. OR AND 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 inches (TAF)
| |
| : b. (Reactor vessel/RCS [PWR] or RPV [BWRJ) level less than (site-specific level). OR
| |
| : 3. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be 3. a. RPV water level cannot be monitored for > 30 minutes monitored for 30 minutes or longer.
| |
| AND AND
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| : b. Core uncovery is indicated by ANY of the following:
| |
| " Table C3 indications of a sufficient magnitude to indicate core
| |
| " (Site-specific radiation monitor) reading greater than (site- uncovery.
| |
| specific value)
| |
| OR
| |
| " Erratic source range monitor indication [PWR]
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/hr.
| |
| " UNPLANNED increase in (site-specific sump and/or tank levels) of sufficient magnitude to indicate core uncovery
| |
| " (Other site-specific indications) Table C3 Indications of RCS Leakage
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| * UNPLANNED Torus level rise*
| |
| * UNPLANNED vessel make up rate rise
| |
| * Observation of leakage or inventory loss
| |
| *Rise in level is attributed to a loss of RPV inventory.
| |
| Page 49 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification CA1 CA6 Initiating Condition: ALERT Initiating Condition: D: No Change M Difference 1 Deviation Loss of (reactor vessel/RCS [PWRI or RPV [BWR]) inventory Loss of RPV inventory 1) Listed site specific levels, and sumps and tanks to ensure timely Operating Mode Applicability: Operating Mode Applicability: classification.
| |
| Cold Shutdown, Refueling 4,5 Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Alert promptly upon Note:
| |
| determining that 15 minutes time has been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be
| |
| : 1. Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as exceeded.
| |
| indicated by level less than (site-specific level).
| |
| : 2. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer 1. Loss of RPV inventory as indicated by level < - 59inches.
| |
| AND OR
| |
| : b. UNPLANNED increase in (site-specific sump and/or tank) levels due 2. a. RPV water level cannot be monitored for_> 15 minutes.
| |
| to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. AND
| |
| : b. Loss of RPV inventory per Table C3 indications.
| |
| Table C3 Indications of RCS Leakage
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| * UNPLANNED Torus level rise*
| |
| * UNPLANNED vessel make up rate rise
| |
| * Observation of leakage or inventory loss
| |
| *Rise in level is attributed to a loss of RPV inventory.
| |
| Page 50 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification Initiating Condition: UNUSUAL EVENT Cul Initiating Condition:
| |
| CU6 H No Change Difference Deviation UNPLANNED loss of (reactor vessel/RCS [PWRI or RPV [BWR]) inventory for UNPLANNED loss of reactor vessel / RCS inventory for 15 minutes or longer. 1) Described "a required lower limit" as a procedurally established lower limit, 15 minutes or longer. and Operating Mode Applicability:
| |
| Operating Mode Applicability: listed site specific sumps and tanks to ensure timely classification.
| |
| 4,5 Cold Shutdown, Refueling Emergency Action Level (EAL):
| |
| Example Emergency Action Levels: (1 or 2)
| |
| Note:
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be
| |
| * The Emergency Director should declare the event promptly upon exceeded. determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| : 1. UNPLANNED loss of reactor coolant results in (reactor vessel/RCS
| |
| [PWRj or RPV [BWRI) level less than a required lower limit for 15 minutes or longer. 1. UNPLANNED loss of reactor coolant results in the inability to restore and
| |
| : 2. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be maintain RPV water level above the procedurally established lower limit monitored. for > 15 minutes.
| |
| AND OR
| |
| : b. UNPLANNED increase in (site-specific sump and/or tank) levels. 2. a. RPV water level cannot be monitored AND
| |
| : b. Loss of RPV inventory per Table C3 indications.
| |
| Table C3 Indications of RCS Leakage
| |
| * UNPLANNED floor or equipment sump level rise*
| |
| * UNPLANNED Torus level rise*
| |
| * UNPLANNED vessel make up rate rise
| |
| * Observation of leakage or inventory loss
| |
| *Rise in level is attributed to a loss of RPV inventory.
| |
| Page 51 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HG1 HG1 Initiating Condition: GENERAL EMERGENCY Initiating Condition: D No Change M Difference F- Deviation HOSTILE ACTION resulting in loss of physical control of the facility. HOSTILE ACTION resulting in loss of physical control of the facility. 1) List site security shift supervision as Security Force.
| |
| : 2) Added descriptors to better explain each safety function and allow for a Operating Mode Applicability: Operating Mode Applicability: timely classification.
| |
| All 1,2,3,4,5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| : 1. a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-specific security shift 1. A notification from the Security Force that a HOSTILE ACTION is occurring or has supervision). occurred within the PROTECTED AREA.
| |
| AND AND
| |
| : 2. a. ANY Table H1 safety function cannot be controlled or maintained.
| |
| : b. EITHER of the following: OR
| |
| : b. Damage to spent fuel has occurred or is IMMINENT
| |
| : 1. ANY of the following safety functions cannot be controlled or maintained.
| |
| Table H1 Safety Functions
| |
| * Reactivity control " Reactivity Control
| |
| * Core cooling [PWR] I RPV water level [BWR] (ability to shut down the reactor and keep it shutdown)
| |
| * RCS heat removal " RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| OR
| |
| : 2. Damage to spent fuel has occurred or is IMMINENT Page 52 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HSI HSI Initiating Condition: SITE AREA EMERGENCY Initiating Condition: No Change D Difference F Deviation HOSTILE ACTION within the Protected Area. HOSTILE ACTION within the Protected Area. 1) List site security shift supervision as Security Force.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| A HOSITLE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site-security shift supervision). A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.
| |
| Page 53 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HAI HAI Initiating Condition: ALERT Initiating Condition: No Change FIDifference i1 Deviation HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack 1) List site security shift supervision as Security Force.
| |
| threat within 30 minutes. threat within 30 minutes.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: (1 or 2) Emergency Action Level (EAL):
| |
| : 1. A HOSTILE ACTION is occurring or has occurred within the OWNER
| |
| : 1. A validated notification from NRC of an aircraft attack threat < 30 minutes CONTROLLED AREA qs reported by the (site-specific security shift from the site.
| |
| supervision).
| |
| : 2. A validated notification from NRC of an aircraft attack threat within 30 OR minutes of the site.
| |
| : 2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.
| |
| Page 54 of 66
| |
| | |
| NEI 99-01 Rev 6 j Proposed EAL Justification HUI HUI Initiating Condition: UNUSUAL EVENT Confirmed SECURITY CONDITION or threat.
| |
| Initiating Condition: FD1 No Change Fq Difference F Deviation Confirmed SECURITY CONDITION or threat. 1) List site security shift supervision as Security Force.
| |
| Operating Mode Applicability: 2) Further described credible security threat through listing a site specific Operating Mode Applicability: procedure.
| |
| All 1,2, 3, 4, 5, D Example Emergency Action Levels: (1 or 2 or 3)
| |
| Emergency Action Level (EAL):
| |
| : 1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision). 1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.
| |
| : 2. Notification of a credible security threat directed at the site. OR
| |
| : 2. A validated notification from the NRC providing information of an aircraft
| |
| : 3. A validated notification from the NRC providing information of an aircraft threat.
| |
| threat.
| |
| OR
| |
| : 3. Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.
| |
| Page 55 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL [ Justification Initiating Condition: SITE AREA EMERGENCY HS6 Initiating Condition:
| |
| HS2 H* No Change Difference Deviation Inability to control a key safety function from outside the Control Room. Inability to control a key safety function from outside the Control Room. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would Operating Mode Applicability: Operating Mode Applicability: enter should such an event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate All 1, 2, 3,4, 5, D shutdown were to be listed,
| |
| : 2) Added descriptors to better explain each safety function and allow for a Example Emergency Action Levels: (1 and 2) Emergency Action Level (EAL): timely classification.
| |
| Note: The Emergency Director should declare the Site Area Emergency Note:
| |
| promptly upon determining that (site-specific number of minutes) has 3) Changed "An event" to" A Control Room evacuation" to remove confusion been exceeded, or will likely be exceeded. The Emergency Director should declare the event promptly upon determining that if partial plant control was transferred to outside the control room with the the applicable time has been exceeded, or will likely be exceeded. control room still manned, due to testing or equipment failure.
| |
| : 1. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). 1 A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per
| |
| : 2. Control of ANY of the following key safety functions is not reestablished
| |
| * QCARP 0050-01, SB-1-llnjection with SSMP and Bringing the Unit to within (site-specific number of minutes). Cold Shutdown OR Reactivity control
| |
| * QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to
| |
| * Core cooling [PWR] / RPV water level [BWR] Cold Shutdown RCS heat removal OR
| |
| * ODA 0010-05, Plant Operation with the Control Room Inaccessible AND
| |
| : 2. Control of ANY Table H1 key safety function is not reestablished in < 15 minutes.
| |
| Table H1 Safety Functions
| |
| * Reactivity Control (ability to shut down the reactor and keep it shutdown)
| |
| * RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| Page 56 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL Justification HA6 HA2 Initiating Condition: ALERT Initiating Condition: D No Change F Difference FIDeviation Control Room evacuation resulting in transfer of plant control to alternate locations. Control Room evacuation resulting in transfer of plant control to alternate locations. 1) EAL uses the site specific Control Room evacuation procedure to effectively list all of the alternate locations, panels, and stations requested by the developer notes. This would be the procedure the Control Room would enter should such an Operating Mode Applicability: Operating Mode Applicability: event occur, this allows for greater clarity as to when this EAL would apply than if each panel and station used in alternate shutdown were to be listed, All 1, 2, 3,4, 5, D
| |
| : 2) Changed "An event" to" A Control Room evacuation" to remove confusion if partial plant control was transferred to outside the control room with the control Example Emergency Action Levels: Emergency Action Level (EAL): room still manned, due to testing or equipment failure.
| |
| An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations). A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:
| |
| " QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| " QOA 0010-05, Plant Operation with the Control Room Inaccessible Page 57 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL j Justification HU4 HU3 Initiating Condition: UNUSUAL EVENT Initiating Condition: No Change EIDifference F Deviation FIRE potentially degrading the level of safety of the plant.
| |
| FIRE potentially degrading the level of safety of the plant. 1) Listed site specific list of plant rooms or areas that contain SAFETY SYSTEM equipment to ensure timely classification.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| : 2) Added additional note to easily direct the operator to escalation ICs All 1, 2, 3,4, 5, D Example Emergency Action Levels: (1 or 2 or 3 or 4)
| |
| Emergency Action Level (EAL):
| |
| Note: The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been Note:
| |
| exceeded, or will likely be exceeded.
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| : 1. a. A FIRE is NOT extinguished within 15-minutes of ANY of the
| |
| * Escalation of the emergency classification level would be via IC CA2 or MA5 following FIRE detection indications:
| |
| : 1. A FIRE in ANY Table H2 area is not extinguished in < 15-minutes of ANY of the
| |
| * Report from the field (i.e., visual observation)
| |
| * Receipt of multiple (more than 1) fire alarms or indications following FIRE detection indications:
| |
| * Field verification of a single fire alarm
| |
| * Report from the field (i.e., visual observation)
| |
| AND
| |
| * Receipt of multiple (more than 1) fire alarms or indications
| |
| : b. The FIRE is located within ANY of the following plant rooms or
| |
| * Field verification of a single fire alarm areas: Table H2 Vital Areas (site-specific list of plant rooms or areas)
| |
| * Reactor Building (when inerted the Drywell is exempt)
| |
| * Main Control Room Envelope
| |
| : 2. a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).
| |
| * Unit and Shared Emergency Diesel Generator Rooms AND
| |
| * 4KV Switchgear Area
| |
| : b. The FIRE is located within ANY of the following plant rooms or
| |
| * Battery Rooms areas:
| |
| * RHR Service Water Vaults (site-specific list of plant rooms or areas)
| |
| * Turbine Building Cable Tunnel AND
| |
| * Cribhouse
| |
| : c. The existence of a FIRE is riot verified within 30-minutes of alarm OR receipt.
| |
| : 2. a. Receipt of a single fire alarm in ANY Table H2 area (i.e., no other indications of a FIRE).
| |
| : 3. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the AND plant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. b. The existence of a FIRE is not verified in < 30 minutes of alarm receipt.
| |
| OR
| |
| : 4. A FIRE within the plant or ISFSI [for plants with an ISFSI outside the 3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-plant Protected Area] PROTECTED AREA that requires firefighting minutes of the initial report, alarm or indication.
| |
| support by an offsite fire response agency to extinguish. OR
| |
| : 4. A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
| |
| Page 58 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HU2 Initiating Condition: UNUSUAL EVENT Initiating Condition:
| |
| HU4 D No Change
| |
| * Difference F Deviation Seismic event greater than OBE levels. 1) Used Alternate developer notes allowed wording since specific Control Room Seismic event greater than OBE levels.
| |
| indication of a seismic event > OBE is not available.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability: 2) Added additional compensatory thresholds based on the guidance provided in US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear All Power Plant Operator Earthquake Actions Appendix A, also added a 15 min time 1,2, 3,4, 5, D limit to ensure timely classification when these additional thresholds may not be Example Emergency Action Levels: Emergency Action Level (EAL): readily available.
| |
| Note: Escalation of the emergency classification level would be via IC CA2 or Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: MA5 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
| |
| : a. (site-specific indication that a seismic event met or exceeded OBE limits) For emergency classification if EAL 2.b is not able to be confirmed, then will determine classification within 15 mins of the event the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in 1 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.
| |
| Page 59 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification
| |
| -F -I H HA5 A5 Initiating Condition: ALERT Initiating Condition: D No Change ] Difference D eviation Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or 1) Listed plant specific rooms and areas with entry related mode Gaseous release impeding access to equipment necessary for normal shutdown. applicability to ensure timely classification.
| |
| plant operations, cooldown or shutdown.
| |
| : 2) The Operating Mode Applicability of this EAL has been revised from All Operating Mode Applicability: Modes to modes 3 and 4 due to the mode applicability of the areas of Operating Mode Applicability: concern in Table H-3.
| |
| 3,4 All Emergency Action Level (EAL):
| |
| Example Emergency Action Levels:
| |
| Note:
| |
| Note: If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no If the equipment in the listed room or area was already inoperable, or out of service, before emergency classification is warranted. the event occurred, then no emergency classification is warranted.
| |
| : 1. a. Release of a toxic, corrosive, asphyxiant or flammable gas
| |
| : 1. Release of a toxic, corrosive, asphyxiant or flammable gas in a Table H3 area.
| |
| into any of the following plant rooms or areas:
| |
| (site-specific list of plant rooms or areas with entry-related Table H3 mode applicability identified) Areas with Entry Related Mode Applicability Entry Related AND Area Unit Mode Applicability
| |
| : b. Entry into the room or area is prohibited or impeded. Reactor Building
| |
| " First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| " First Floor South Wall 2
| |
| " Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 AND
| |
| : 2. Entry into the room or area is prohibited or impeded.
| |
| Page 60 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HU3 HU6 Initiating Condition: UNUSUAL EVENT Initiating Condition: D No Change FIDifference F- Deviation Hazardous Event Hazardous Event 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 Operating Mode Applicability: EAL UE values.
| |
| Operating Mode Applicability:
| |
| All 2) Changed the word "needed" to "required by Technical Specifications" in the EAL 1,2, 3,4, 5, D to be consistent with terminology used by operators and minimize confusion.
| |
| Example Emergency Action Levels: (1 or 2 or 3 or 4)
| |
| Emergency Action Level (EAL): 3) Added additional note to easily direct the operator to escalation ICs Note: EAL #3 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
| |
| Note:
| |
| I. A tornado strike within the PROTECTED AREA.
| |
| * EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
| |
| : 2. Internal room or area flooding of a magnitude sufficient to require manual
| |
| * Escalation of the emergency classification level would be via IC CA2 or or automatic electrical isolation of a SAFETY SYSTEM component needed MA5 for the current operating mode.
| |
| : 1. Tornado strike within the PROTECTED AREA.
| |
| : 3. Movement of personnel within the PROTECTED AREA is impeded due to OR an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release). 2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by
| |
| : 4. A hazardous event that results in on-site conditions sufficient to prohibit Technical Specifications for the current operating mode.
| |
| the plant staff from accessing the site via personal vehicles. OR
| |
| : 5. (Site-specific list of natural or technological hazard events) 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 River level, as indicated by EITHER:
| |
| : a. High river level > 594 ft.
| |
| OR
| |
| : b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam #14 has failed.
| |
| Page 61 of 66
| |
| | |
| NEI 99-01 Rev 6 1 Proposed EAL I Justification HG7 HG7 Initiating Condition: GENERAL EMERGENCY Initiating Condition: W No Change EIDifference F Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. declaration of a General Emergency.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1,2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT Other conditions exist which in the judgment of the Emergency Director indicate substantial core degradation or melting with potential for loss of containment that events are in progress or have occurred which involve actual or IMMINENT integrity or HOSTILE ACTION that results in an actual loss of physical control of substantial core degradation or melting with potential for loss of containment the facility. Releases can be reasonably expected to exceed EPA Protective Action integrity or HOSTILE ACTION that results in an actual loss of physical control of Guideline exposure levels off-site for more than the immediate site area. the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area..
| |
| Page 62 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: SITE AREA EMERGENCY Initiating Condition: LSi No Change [JDifference Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. declaration of a Site Area Emergency.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Other conditions exist which in the judgment of the Emergency Director indicate Other conditions exist which in the judgment of the Emergency Director indicate that major that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION fainu that results in intentional damage or malicious acts; (1) toward site personnel or failures of plant functions needed for protection of the public or HOSTILE ACTION equipment that could lead to the likely failure of or; (2) that prevent effective that results in intentional damage or malicious acts; (1) toward site personnel or access to equipment needed for the protection of the public. Any releases are not equipment that could lead to the likely failure of or; (2) that prevent effective expected to result in exposure levels which exceed EPA Protective Action access to equipment needed for the protection of the public. Any releases are not Guielinetedtoresultin exposure levels be thextebuda, Eexpected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. Guideline exposure levels beyond the site boundary.
| |
| Page 63 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification Initiating Condition: ALERT Initiating Condition: Lhi No Change [J Difference Deviation Other conditions exist which in the judgment of the Emergency Director warrant Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. declaration of an Alert.
| |
| Operating Mode Applicability: Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels: Emergency Action Level (EAL):
| |
| Other conditions exist which in the judgment of the Emergency Director indicate Other conditions exist which in the judgment of the Emergency Director indicate that events degradation substantial are in progress or have of the level occurred of safety which involve of the plant or an actual orevent a security potential that that events are in progress or have occurred which involve an actual or potential substanvolvdesgpradatn lfe threatenln rkof safetyohe peonn ordamagetosityev 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 siteinovspballfetrtnngiktoiepronlordmetoie equipment because of HOSTILE ACTION. Any releases are expected to be involves probable life threatening risk to site personnel or damage to site limited to small fractions of the EPA Protective Action Guideline exposure levels, equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
| |
| Page 64 of 66
| |
| | |
| NEI 99-01 Rev 6 Proposed EAL Justification HU7 HU7 r m m Initiating Condition: UNUSUAL EVENT Initiating Condition: L No Change Difference Deviation Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT. Other conditions existing which in the judgment of the Emergency director warrant declaration of an UNUSUAL EVENT.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| All 1,2, 3,4, 5, D Example Emergency Action Levels:
| |
| Other conditions exist which in the judgment of the Emergency Director indicate Emergency Action Level (EAL):
| |
| that events are in progress or have occurred which indicate a potential Other conditions exist which in the judgment of the Emergency Director indicate degradation of the level of safety of the plant or indicate a security threat to facility that events are in progress or have occurred which indicate a potential protection has been initiated. No releases of radioactive material requiring offsite degradation of the level of safety of the plant or indicate a security threat to facility response or monitoring are expected unless further degradation of safety systems protection has been initiated. No releases of radioactive material requiring offsite occurs. response or monitoring are expected unless further degradation of safety systems occurs.
| |
| Page 65 of 66
| |
| | |
| NEI 99-01 Rev 6 [ Proposed EAL [ Justification E-HU1 Initiating Condition: UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY.
| |
| Initiating Condition:
| |
| E-HUI W-] No Change 1 Difference i Deviation Damage to a loaded cask CONFINEMENT BOUNDARY. 1) Listed 2x the site specific cask specific allowable radiation level as per Certificate of Compliance No. 1014 Appendix A Section 5.7.
| |
| Operating Mode Applicability:
| |
| Operating Mode Applicability:
| |
| All 1, 2, 3,4, 5, D Example Emergency Action Levels:
| |
| Emergency Action Level (EAL):
| |
| Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-technical specification allowable radiation level) on the surface of the spent fuel contact radiation reading:
| |
| cask.
| |
| * > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR
| |
| * > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts.
| |
| Page 66 of 66
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 Initiating Condition:
| |
| Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRem TEDE or 5000 mRem thyroid CDE.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| * Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond 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.
| |
| Month 20XX QC 3-33 EP-AA-1006 (Revision XX)
| |
| | |
| (1"2ri ifina Annov I=:vgm-In N"Aaln-r flu .~.a~
| |
| ~- ri Ci*ic~c
| |
| -
| |
| * t*~o Anncv 5-s. .. . n ~ . ~. . . . - ~. - ~A.
| |
| RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 (cont)
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG1
| |
| : 2. EP-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 4. QGA 400 Radioactivity Release Control Month 20XX QC 3-34 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RS1 Initiating Condition:
| |
| Release of gaseous radioactivity resulting in offsite dose greater than 100 mRem TEDE or 500 mRem thyroid CDE.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| 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.
| |
| Month 20XX QC 3-35 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RSI (cont)
| |
| 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-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-36 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| * Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
| |
| : a. > 10 mRem TEDE OR
| |
| : b. > 50 mRem CDE Thyroid OR Month 20XX QC 3-37 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
| |
| : a. 10 mRem TEDE for 60 minutes of exposure OR
| |
| : b. 50 mRem CDE Thyroid for 60 minutes of exposure OR
| |
| : 4. Field survey results at or beyond the site boundary indicate EITHER:
| |
| : a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for > 60 minutes.
| |
| OR
| |
| : b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Basis:
| |
| This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of 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.
| |
| Escalation of the emergency classification level would be via IC RS1.
| |
| Month 20XX QC 3-38 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA1
| |
| : 2. CY-QC-120-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. QOP 2000-24, Discharging to the River from the River Discharge Tank using the Waste Surge Pump
| |
| : 6. QOP 2000-25, Discharging to the River from the River Discharge Tank using the River Discharge Pump
| |
| : 7. CY-QC-1 20-729, Liquid Effluent Alarm Setpoints
| |
| : 8. QCOA 1700-02, High Radiation detected on Eberline Radiation Monitoring System
| |
| : 9. QCAN 912-5 C-6, Radwaste High Rad.
| |
| : 10. QCAN 901(2)-3-G-1, Liquid Process Rad. Monitor High Radiation
| |
| : 11. CY-QC-1 20-735, Main Chimney & Reactor Vent Noble Gas Release Rate Action Levels
| |
| : 12. QCOA 1700-01, Abnormal Chimney Radiation
| |
| : 13. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 14. QGA 400 Radioactivity Release Control Month 20XX QC 3-39 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 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, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
| |
| : 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for > 60 minutes.
| |
| " Radwaste Effluent Monitor 1/2-1799-01 OR
| |
| " Discharge Permit specified monitor OR
| |
| : 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
| |
| Basis:
| |
| This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
| |
| Month 20XX QC 3-40 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RUl (cont)
| |
| Basis (cont):
| |
| 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.
| |
| EAL #1 Basis This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
| |
| The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.
| |
| EAL #2 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous effluent pathways.
| |
| EAL #3 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 RA1.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU1
| |
| : 2. CY-QC-120-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-41 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, 6, 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.
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG2 Month 20XX QC 3-42 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex OuadCites AnexExelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS2 Initiating Condition:
| |
| Spent fuel pool level at (site-specific Level 3 description).
| |
| Operating Mode Applicability:
| |
| 1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, AS2 Month 20XX QC 3-43 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 Initiating Condition:
| |
| Significant lowering of water level above, or damage to, irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1, 2, 3,4, 5, D Emergency Action Level (EAL):
| |
| : 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.
| |
| OR
| |
| : 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R1 Radiation Monitor reading >1000 mRem/hr.
| |
| OR
| |
| : 3. Lowering of spent fuel pool level to (site specific Level 2 value).
| |
| Table R1 Fuel Handling Incident Radiation Monitors 0 1(2) 1705-16A Fuel Pool Rad Monitor
| |
| * 1(2) 1705-16B Fuel Pool Rad Monitor Basis:
| |
| REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.
| |
| IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
| |
| CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-44 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 (cont)
| |
| Basis (cont):
| |
| 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 an 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 Ror C ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA2
| |
| : 2. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity
| |
| : 3. QCAN 901(2)-3 B-1 Refuel Floor Hi Radiation
| |
| : 4. QCAN 901(2)-3 G-16/H-16 Fuel Pool Channel A/B Hi Radiation
| |
| : 5. QCIS 1700-07 Reactor Building Ventilation and Fuel Pool Radiation Monitoring Calibration and Functional Test Month 20XX QC 3-45 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 Initiating Condition:
| |
| UNPLANNED loss of water level above irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| : 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
| |
| " Refueling Cavity water level < 282 in. (Upper Wide range simulated signal).
| |
| OR
| |
| " Spent Fuel Pool water level < 19 ft. above the fuel (<- 4 ft. indicated level).
| |
| OR
| |
| " Indication or report of a drop in water level in the REFUELING PATHWAY.
| |
| AND
| |
| : b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R1.
| |
| Table R1 Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Mon
| |
| * 1(2) 1705-16B Fuel Pool Rad Mon Month 20XX QC 3-46 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 (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.
| |
| 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 an 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU2
| |
| : 2. QCOP 0201-13 Reactor Level Upper Wide Range Reference Leg Extension Use and Control
| |
| : 3. Technical Specifications 3.7.8 Spent Fuel Storage Pool Water Level
| |
| : 4. Technical Specifications 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel
| |
| : 5. QCAN 901(2)-4 B-24 FUEL POOL STORAGE HI/LO LEVEL
| |
| : 6. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity Month 20XX QC 3-47 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Exelon Nnncl*ar Nuclea~r Duasd Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RA3 Initiating Condition:
| |
| Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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 R2 areas:
| |
| Table R2 Areas Requiring Continuous Occupancy
| |
| " Main Control Room (Unit 1 ARM Station #22)
| |
| * Central Alarm Station - (by survey)
| |
| OR
| |
| : 2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:
| |
| Table R3 Arcas with ENtr Related -, Mode Appliabiity A~eEntry Rclatcd Mode
| |
| __Appllr~abillity.
| |
| Reactor Building* Mode 3 and-4 Turbine Building
| |
| * Mode 3
| |
| *Areas required to establish shutdown GEcoolin Month 20XX QC3-48 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Table R3 Areas with Entry Related Mode Applicability Area Unit Entry Related Mode Applicability Reactor Building
| |
| " First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 Month 20XX QC 3-49 EP-AA-1 006 (Revision XX)
| |
| | |
| FYeInn NHP.I*_*r Ouadr Cities~ AnnAex Fvyinnn Nmriar~I RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.
| |
| This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
| |
| Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.
| |
| For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in 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.
| |
| " 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.).
| |
| Month 20XX QC 3-50 EP-AA-1006 (Revision XX)
| |
| | |
| l:YpInn N.r.l*n r Dua~d Cities Annex FyDinn M..r~iozr RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS RA3 (cont)
| |
| Basis (cont):
| |
| " 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. QCOP 1800-1 Operation of ARM Indicator/Trip Units
| |
| : 3. UFSAR Section 3.2
| |
| : 4. General Arrangement Drawings M-5, 6, 8 and 10 Month 20XX QC 3-51 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU3 Initiating Condition:
| |
| Reactor coolant activity greater than Technical Specification allowable limits.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| : 1. Offgas system radiation monitor HI-HI alarm.
| |
| OR
| |
| : 2. Specific coolant activity > 4.0 uCi/gm Dose equivalent 1-131.
| |
| Basis:
| |
| This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.
| |
| Conditions that cause the specified monitor to alarm that are not related to fuel clad degradation should not result in the declaration of an Unusual Event.
| |
| This EAL addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity.
| |
| An Unusual EVent is only warranted when actual fuel clad damage is the cause of the elevated coolant sample activity (as deteFrmined by laboratory confiFrmation). Fuel cla damage should be assumed to be the cause o)f elevated Reactor Coolant activity unless anoher c ause is known.
| |
| Escalation of the emergency classification level would be via I(s FA1 or the Recognition Category R ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU3
| |
| : 2. Technical Specifications 3.4.6
| |
| : 3. Technical Specifications 3.7.6
| |
| : 4. QCAN 901(2)-3 C-2 OFF GAS HIGH-HIGH RADIATION Month 20XX QC 3-52 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:
| |
| Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the General Emergency classification level each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-53 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:
| |
| Loss or Potential Loss of ANY two barriers.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the Site Area Emergency classification level, each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-54 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:
| |
| ANY Loss or ANY Potential Loss of EITHER Fuel Clad or RCS.
| |
| Operating Mode Applicability:
| |
| 1,2,3 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 Month 20XX QC 3-55 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn N.r.lpar r~uad Cities Annex FYAlnn Nucler~I RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:
| |
| RCS Activity Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Coolant activity > 300 uCi/gm Dose Equivalent 1-131.
| |
| Basis:
| |
| This threshold indicates that RCS radioactivity concentration is greater than 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.
| |
| 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
| |
| : 2. NF-AA-430, Failed Fuel Action Plan Month 20XX QC 3-56 EP-AA-1006 (Revision XX)
| |
| | |
| Duadr Cities~ Annex Exelon Nuceiair RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:
| |
| RPV Water Level 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 > -142 inches (TAF)
| |
| OR
| |
| : 3. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.
| |
| This is identified in the BWROG EOPs/SAMGs 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 Month 20XX QC 3-57 EP-AA-1006 (Revision XX)
| |
| | |
| fl"!arl rifiac Annov IPvnlnn hn~rlaar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)
| |
| Basis (cont):
| |
| RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
| |
| The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-58 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Drywell radiation monitor reading > 6.65 E+02 R/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 300 ý0Ci/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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. Core Damage Assessment Methodology (CDAM)
| |
| Month 20XX QC 3-59 EP-AA-1006 (Revision XX)
| |
| | |
| (1".qd rifloa Annoy I:::YlrIn Miaerla-r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-60 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. RPV water level cannot be restored and maintained > -142 inches (TAF)
| |
| OR
| |
| : 2. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.
| |
| The RPV water level threshold is the same as Fuel Clad 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 Month 20XX QC 3-61 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex OuadCites AnexExelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)
| |
| Basis (cont):
| |
| fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 500-4 RPV Flooding Month 20XX QC 3-62 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:
| |
| Primary Containment Pressure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell pressure >2.5 psig.
| |
| AND
| |
| : 2. Drywell pressure rise is due to RCS leakage.
| |
| Basis:
| |
| The > 2.5 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.
| |
| The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.
| |
| The release of mass from the RCS due to the as-designed/expected operation of any relief valve does not warrant an emergency classification.
| |
| A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| There is no Potential Loss threshold associated with Primary Containment Pressure.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 200 Primary Containment Control Month 20XX QC 3-63 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:
| |
| RCS Leak Rate Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC 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 > QGA 300 Maximum Normal operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300 Maximum Normal 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 on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.
| |
| 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.
| |
| Month 20XX QC 3-64 EP-AA-1006 (Revision XX)
| |
| | |
| n"nrl Mae Annav F~vale~n NMgirInr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis (cont):
| |
| 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 safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the Torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
| |
| Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
| |
| A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
| |
| The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.
| |
| In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.
| |
| An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold
| |
| #1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
| |
| Month 20XX QC 3-65 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. M-13 Main steam piping
| |
| : 3. UFSAR 5.2.5
| |
| : 4. QCOA 0201-01 Rev 016, Increasing Drywell Pressure
| |
| : 5. QOA 900-4 A-17 900-4 A-17 Annunciator
| |
| : 6. QCOS 1600-07 Reactor Coolant Leakage In The Drywell
| |
| : 7. QGA 300 Secondary Containment Control Month 20XX QC 3-66 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:
| |
| Primary Containment radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell radiation monitor reading> I OOR/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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. CaIc. EP-EAL-0611 Month 20XX QC 3-67 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:
| |
| 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.
| |
| 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 Month 20XX QC 3-68 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-69 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn FvAlnn Nuclear NuclerIzr Duad Cities Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:
| |
| Primary Containment 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 > 56 psig and rising.
| |
| OR
| |
| : 4. a. Drywell or torus hydrogen concentration > 6%.
| |
| AND
| |
| : b. Drywell or torus oxygen concentration > 5%.
| |
| OR
| |
| : 5. Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| Loss Threshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
| |
| These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.
| |
| Month 20XX QC 3-70 EP-AA-1006 (Revision XX)
| |
| | |
| n"nd ritia4z Annoy IPvle, n Ih, r~lanr Am i~.rI Citi~~ Ann~v FvaIr~n MmirI~2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)
| |
| Basis (cont):
| |
| Potential Loss Threshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.
| |
| Potential Loss 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. UFSAR Fig. 6.2-16a
| |
| : 3. UFSAR Section 15.6
| |
| : 4. UFSAR 6.2.1.1
| |
| : 5. QGA 200, Primary Containment Control
| |
| : 6. Quad Cities PSTG Section 5, Primary Containment Control Month 20XX QC 3-71 EP-AA-1 006 (Revision XX)
| |
| | |
| ExAlnn Nuclear Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS
| |
| : 1. Drywell radiation monitor reading > 1.55 E+03 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-1 228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20%
| |
| in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. Core Damage Assessment Methodology (CDAM)
| |
| Month 20XX QC 3-72 EP-AA-1006 (Revision XX)
| |
| | |
| Oiiad Cities Annex Exelon Nuclear I=xelon Nucle~ar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:
| |
| Primary Containment Isolation Failure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 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 > QGA 300, Maximum Safe operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300, Maximum Safe operating levels.
| |
| 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, HPCI or RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.
| |
| Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.
| |
| Month 20XX QC 3-73 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Month 20XX QC 3-74 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)
| |
| Basis (cont):
| |
| Main Turbine Control Valve closure) may indicate that a, primary system is discharging into the Reactor Building.
| |
| In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.
| |
| 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. QGA 200 Primary Containment Control
| |
| : 3. QGA 200-5 Hydrogen Control
| |
| : 4. QCOP 1600-13 Post-Accident Venting of the Primary Containment
| |
| : 5. QGA 300 Secondary Containment Control Month 20XX QC 3-75 EP-AA-1006 (Revision XX)
| |
| | |
| A 112 rl Cificn Annov I::v-Inln N, n-I-nr RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-76 EP-AA-1006 (Revision XX)
| |
| | |
| fl"!mfl ('ifiac Annov Pvalewn N~irlanr 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. EITHER of the following:
| |
| : a. Restoration of at least one unit ECCS bus in < 1 hour is not likely.
| |
| OR
| |
| : b. RPV water level cannot be restored and maintained > -166 inches.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.
| |
| This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat 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.
| |
| Month 20XX QC 3-77 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellan Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)
| |
| Basis (cont):
| |
| Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
| |
| The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
| |
| The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. QGA 100 RPV Control Month 20XX QC 3-78 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Nuclear QuadCites AnexExelon 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.
| |
| Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
| |
| Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.
| |
| Month 20XX QC 3-79 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon NuclIear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-80 EP-AA-1006 (Revision XX)
| |
| | |
| rliinfl riflac Annav Pvi:=nn klielanr A. .~ w4Citig~ Annoy Fvoinn Mm me102r RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit Auxiliary Transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers 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 MUl.
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| * A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
| |
| Month 20XX QC 3-81 EP-AA-1006 (Revision XX)
| |
| | |
| Exallon Eeo Nuclear ula Quad Cities Annex RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)
| |
| Basis (cont):
| |
| * A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-82 EP-AA-1006 (Revision XX)
| |
| | |
| 01iad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3 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 unit ECCS busses 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. (e.g. unit cross-tie breakers)
| |
| Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
| |
| Escalation of the emergency classification level would be via IC MAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout Month 20XX QC 3-83 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.
| |
| AND
| |
| : 3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.
| |
| AND
| |
| : 4. ALL AC and Vital DC power sources have been lost for >15 minutes.
| |
| 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 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 EALs are met.
| |
| Month 20XX QC 3-84 EP-AA-1006 (Revision XX)
| |
| | |
| l=x*Inn N mmr.lpnr r3uad Cities Annex Fxelon Nue-lanir RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG8
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. Technical Specifications 3.8.4 and B3.8.4
| |
| : 10. UFSAR Section 8.3.2
| |
| : 11. QOP 6900-02 125 VDC Electrical System
| |
| : 12. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-85 EP-AA-1006 (Revision XX)
| |
| | |
| QUad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2 Initiating Condition:
| |
| Loss of all vital DC power for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| Voltage is < 105 VDC on 125 VDC battery busses #1 and #2 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS8
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-86 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 > 5%.
| |
| AND
| |
| : 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
| |
| AND
| |
| : 3. EITHER of the following conditions exist:
| |
| * RPV water level cannot be restored and maintained > -166 inches.
| |
| OR
| |
| * Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| 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.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Escalation of the emergency classification level would be via IC RG1 or FGI.
| |
| Month 20XX QC 3-87 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex I=x*lon Nuclear OuadCites AnexExelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 200 Primary Containment Control Month 20XX QC 3-88 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear 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:
| |
| 1,2 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 > 5%.
| |
| AND
| |
| : 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.
| |
| Basis:
| |
| This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.
| |
| Month 20XX QC 3-89 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| Month 20XX QC 3-90 EP-AA-1006 (Revision XX)
| |
| | |
| O.,ad Cities Annex Exelnn Nuc:l*r I=x*lon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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
| |
| > 5%.
| |
| 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
| |
| > 5%.
| |
| 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 control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
| |
| Month 20XX QC 3-91 EP-AA-1006 (Revision XX)
| |
| | |
| IFvAlnn MhirnIl~ar RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.
| |
| The plant response to the failure of an automatic or manual reactor 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 control 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.
| |
| Month 20XX QC 3-92 EP-AA-1006 (Revision XX)
| |
| | |
| I*xAInn Nuclear Oijad Cities Annex Exelon NuclerIsr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. Technical Specifications Table 3.3.1.1-1
| |
| : 5. Technical Specification 3.3.1.3
| |
| : 6. Technical Specification Bases 3.3.1.1 and 3.3.1.3 Month 20XX QC 3-93 EP-AA-1 006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| " Reactor Power
| |
| " RPV Water Level
| |
| " RPV Pressure
| |
| " Drywell Pressure
| |
| " Torus Level
| |
| " Torus Temperature AND
| |
| : 2. ANY Table M2 transient in progress.
| |
| Table M2 Significant Transients
| |
| " Turbine Trip
| |
| " Reactor Scram
| |
| " ECCS Activation
| |
| * Recirc. Runback > 25% Reactor Power Change
| |
| * Thermal Power oscillations > 10% Reactor Power Change Month 20XX QC 3-94 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear 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 Month 20XX QC 3-95 EP-AA-1006 (Revision XX)
| |
| | |
| rl"nrl ifina Annoy IPyalen N~rlalr Am
| |
| -- i~id
| |
| - - -Ci*iaQ
| |
| *U - Ann~v F 5~*
| |
| * S ~fl Fv~InnS M.urIa~r U- ~
| |
| RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| * Reactor Power
| |
| * RPV Water Level
| |
| * RPV Pressure
| |
| * Drywell Pressure
| |
| * Torus Level
| |
| * Torus Temperature Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
| |
| As used in this EAL, an "inability to monitor" means that values for 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.
| |
| Month 20XX QC 3-96 EP-AA-1006 (Revision XX)
| |
| | |
| Mad Cities Annex Nuclear OuadCites AnexExelon 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 Month 20XX QC 3-97 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 Initiating Condition:
| |
| 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.
| |
| 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 Month 20XX QC 3-98 EP-AA-1006 (Revision XX)
| |
| | |
| n"nrl rifina Annoy l::vg* In Ni,,el-anr A u12r4
| |
| -- - - -Citi~~c
| |
| *~ ~Annav
| |
| *~ * ~fl Fv~Inn
| |
| ~ . . - *vIa~r
| |
| * ~. . Nh ~
| |
| RECOGNITION CATEGORY SYSTEM MALFUNCTIONS events may require a post-event inspection to determine if the attributes of an explosion are present.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. These are typically systems classified as safety-related.
| |
| Month 20XX QC 3-99 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exallon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| This EAL #2a 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.
| |
| This 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 is not 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 SA9 Month 20XX QC 3-100 EP-AA-1006 (Revision XX)
| |
| | |
| Dijad Cities Annex OuadCitesRECOGNITION AnexExelon CATEGORY Nuclear SYSTEM MALFUNCTIONS 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.
| |
| 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 Month 20XX QC 3-101 EP-AA-1006 (Revision XX)
| |
| | |
| Omiad Cities Annex Exelon Nuclear Exelon Nuclea~r RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)
| |
| Basis (cont):
| |
| 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 Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| 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. QCOS 1600-07 Reactor Coolant Leakage in the Drywell
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QCOA 0201-01 Increasing Drywell Pressure Month 20XX QC 3-102 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Initiating Condition:
| |
| Loss of all On-site or Off-site communications capabilities.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency 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 Radio X Plant Page X Sound Powered Phones X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite 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.).
| |
| Month 20XX QC 3-103 EP-AA-1 006 (Revision XX)
| |
| | |
| I::xelnn Nuclear O.,ad Cities Annex Exellnn NuclerIzr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis This EAL addresses a total loss of the communications methods used in support of routine plant operations.
| |
| EAL #2 Basis This EAL 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) Form.
| |
| EAL #3 Basis This EAL 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-104 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. Loss of ALL offsite AC power to unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| 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.
| |
| Month 20XX QC 3-105 EP-AA-1006 (Revision XX)
| |
| | |
| I=xplnn NHeI*_*r Quad Cities Annex Exellnn NuclerIsr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA2
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "ACTURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-106 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS cul Initiating Condition:
| |
| Loss of all but one AC power source to emergency buses for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. AC power capability to unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit auxiliary transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers 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.
| |
| Month 20XX QC 3-107 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CUl (cont)
| |
| Basis (cont):
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| * 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 CAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6 CU2
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-108 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:
| |
| Hazardous event affecting SAFETY SYSTEM required for the current operating mode.
| |
| Operating Mode Applicability:
| |
| 4,5 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.
| |
| 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 Month 20XX QC 3-109 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex I=xplnn N HP.Ip_:* r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS events may require a post-event inspection to determine if the attributes of an explosion are present.
| |
| Month 20XX QC 3-110 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA6 Month 20XX QC 3-111 EP-AA-1 006 (Revision XX)
| |
| | |
| FY*_lnn Nnnr.l*r flm,21 CitiA;s Annex ExpInn NMipar~I RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:
| |
| Loss of Vital DC power 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.
| |
| Voltage is < 105 VDC on required 125 VDC battery busses #1 and #2 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 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.
| |
| Month 20XX QC 3-112 EP-AA-1006 (Revision XX)
| |
| | |
| Omar] ritioc Annoy FI:vlrln M,,ril r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CU4
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-113 EP-AA-1006 (Revision XX)
| |
| | |
| (3"2rl Mae Annoy I:volenn nMmr-l-r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:
| |
| Loss of all onsite or offsite communications capabilities.
| |
| Operating Mode Applicability:
| |
| 4,5, 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 C1 Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X Sound Pewcred Phones X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite communications capabilities.
| |
| While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Outside 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.
| |
| Month 20XX QC 3-114 EP-AA-1 006 (Revision XX)
| |
| | |
| Mind ritioc Annoy Nig-lanr P-mairnnM 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-115 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 Initiating Condition:
| |
| Inability to maintain the plant in cold shutdown.
| |
| 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-.
| |
| " 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 > 212OF duc to loss of decay heat removal for > Table C2 duration.
| |
| Table C2 RCS Heat-up Duration Thresholds RCS Containment Closure Heat-up Status Status Duration Intact Not Applicable 60 minutes*
| |
| Not Intact Established 20 minutes*
| |
| 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-te loss of decay heat removal.
| |
| 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.
| |
| Month 20XX QC 3-116 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS 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.).
| |
| Month 20XX QC 3-117 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Month 20XX QC 3-118 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. OU-AA-103 Shutdown Safety
| |
| : 6. QCOA 1000-02 Loss of Shutdown Cooling
| |
| : 7. QGA 100 RPV Control
| |
| : 8. QGA 100 RPV Control Detail A
| |
| : 9. QCGP 1-1 Normal Unit Startup
| |
| : 10. QCIS 0600-01 Unit One Division 1 Reactor Pressure 0 to 1200 psig Indication Calibration Month 20XX QC 3-119 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Ouad Cties nnex xelon Nuc~lear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 Initiating Condition:
| |
| UNPLANNED rise in RCS temperature 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.
| |
| " 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 removal.
| |
| 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.).
| |
| A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
| |
| Month 20XX QC 3-120 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn N Hc.l*_*r Duad Cities~ Annex Fxellnn Nui~clear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 (cont)
| |
| Basis (cont):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, CU3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. QGA 100, RPV Control
| |
| : 4. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 5. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 6. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge
| |
| : 7. QCOA 1000-02 Loss of Shutdown Cooling Month 20XX QC 3-121 EP-AA-1 006 (Revision XX)
| |
| | |
| Exelon Nuclear Eeo ula Duadr Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Initiating Condition:
| |
| Loss of RPV inventory affecting fuel clad integrity with containment challenged.
| |
| 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 a. RPV water level < -142 inches (TAF) for > 30 minutes.
| |
| AND
| |
| : b. Any Containment Challenge Indication (Table C4)
| |
| OR
| |
| : 2. a. RPV water level unknewn 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
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-122 EP-AA-1006 (Revision XX)
| |
| | |
| I=xelnn Exelnn Nucle_*r Nucler~I Omiad Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| Table C4 Containment Challenge Indications
| |
| " Primary Containment Hydrogen Concentration > 6% and Oxygen > 5%
| |
| " UNPLANNED rise in containment pressure
| |
| , CONTAINMENT CLOSURE not established*
| |
| " ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.
| |
| * if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.
| |
| 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.
| |
| Month 20XX QC 3-123 EP-AA-1006 (Revision XX)
| |
| | |
| I=xAInn N, m.l*_*r Dua~d Cities Annex FvAinn Nur~ipar RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis (cont):
| |
| 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-124 EP-AA-1006 (Revision XX)
| |
| | |
| tl"nei rifine Annoy I-on-Itn M,,l"A-r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CG1
| |
| : 2. QGA 100, RPV Control
| |
| : 3. Technical Specifications 3.3.1
| |
| : 4. Technical Specifications 3.6.1.1
| |
| : 5. Technical Specifications 3.6.4.1
| |
| : 6. QGA-200-5, Hydrogen Control
| |
| : 7. UFSAR 6.2.1.1
| |
| : 8. QGA 300 Secondary Containment Control
| |
| : 9. EP-EAL-0501, Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Month 20XX QC 3-125 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 Initiating Condition:
| |
| Loss of RPV inventory affecting core decay heat removal capability.
| |
| 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. With CONTAINMENT CLOSURE not established, RPV water level
| |
| < -65 inches OR
| |
| : 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 inches (TAF).
| |
| OR
| |
| : 3. a. RPV water level cannot be monitoredu'nknown 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
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-126 EP-AA-1006 (Revision XX)
| |
| | |
| I=xe.lnn N Hc.lp_*r Quad Cities Annex Exeiln NIuiclar RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 (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.
| |
| 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.b and 2.b 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 tank levels. Sump and/or tank 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 RG1.
| |
| Month 20XX QC 3-127 EP-AA-1006 (Revision XX)
| |
| | |
| Oijad Cities Annex ExeIon Nuclear I=x*lon Nucler!I RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CS1
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. QGA 100, RPV Control
| |
| : 6. Technical Specifications 3.3.1
| |
| : 7. Technical Specifications Table 3.3.3.1-1
| |
| : 8. Technical Specifications 3.3.5.1
| |
| : 9. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 10. Technical Specifications 3.4.4
| |
| : 11. UFSAR 5.2.5
| |
| : 12. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 13. QOA 900-4 A-17, Annuciator Response Month 20XX QC 3-128 EP-AA-1006 (Revision XX)
| |
| | |
| IFxAlnn NucleIasr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:
| |
| Loss of RPV 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 < - 59 inches.
| |
| OR
| |
| : 2. a. RPV water level cannot be monitored unkwn-;,for > 15 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 in 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 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.
| |
| Month 20XX QC 3-129 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis A lowering of water level below -59 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 decrease in water level will lead to core uncovery.
| |
| 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). An 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 tank levels. Sump and/or tank 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 Specifications 3.3.5.1
| |
| : 3. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 4. Technical Specifications 3.4.4
| |
| : 5. UFSAR 5.2.5
| |
| : 6. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 7. QOA 900-4 A-17, Annuciator Response
| |
| : 8. QGA 100, RPV Control
| |
| : 9. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 10. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 11. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-130 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 Initiating Condition:
| |
| UNPLANNED loss of RPV 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 water level to above the procedurally established lower limit for > 15 minutes.
| |
| OR
| |
| : 2. a. RPV water level cannot be monitoredunknown 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 in RPV inventory Month 20XX QC 3-131 EP-AA-1006 (Revision XX)
| |
| | |
| rl"nd Mac Annoy I~volrln Nl,,Aalr A. i~ ri Ci$i~ Anng~v Fvalnn NImmrI~2r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (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 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.
| |
| 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 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 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-132 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev. 6 CU1
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QGA 100, RPV Control
| |
| : 6. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 7. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 8. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-133 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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, 5, 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 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)
| |
| * RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-134 EP-AA-1006 (Revision XX)
| |
| | |
| Oiiad Cities Annex Exelon NuceaIzr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 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
| |
| : 2. Station Security Plan - Appendix C Month 20XX QC 3-135 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Ou~dCitesRECOGNITION AnexExelon CATEGORY Nuclear HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:
| |
| HOSTILE ACTION within the PROTECTED AREA.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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].
| |
| Month 20XX QC 3-136 EP-AA-1006 (Revision XX)
| |
| | |
| tliinrl r-iflac Annov :valnn kJlrlAa-r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 (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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HS1
| |
| : 3. Station Security Plan - Appendix C Month 20XX QC 3-137 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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.
| |
| Month 20XX QC 3-138 EP-AA-1006 (Revision XX)
| |
| | |
| l:Y*Inn Ninl*_*r Aiisad Citiecs Annex FvaInn N~iJpmrIz RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (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 QCOA 0010-20, Security Event.
| |
| EAL #2 Basis Is 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 HSI.
| |
| Month 20XX QC 3-139 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HA1
| |
| : 2. Station Security Plan - Appendix C
| |
| : 3. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-140 EP-AA-1 006 (Revision XX)
| |
| | |
| Fxelnn NHP.IeRr Oijad !ties Annex FYAlnn Nucler~I RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:
| |
| Confirmed SECURITY CONDITION or threat.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| Basis:
| |
| 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.
| |
| Month 20XX QC 3-141 EP-AA-1006 (Revision XX)
| |
| | |
| l=xnlnn Nucl*nr Quad Cities Annex ExelIn Nucrileazr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HUI (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-101-132.
| |
| 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 QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| 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, HU1
| |
| : 2. SY-AA-101-132, Security Assessment and Response to Unusual Activities
| |
| : 3. Station Security Plan - Appendix C
| |
| : 4. NRC Safeguards Advisory 10/6/01
| |
| : 5. Letter from Mr. B. A. Boger (NRC) to Ms. Lynette Hendricks (NEI) dated 2/4/02
| |
| : 6. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-142 EP-AA-1006 (Revision XX)
| |
| | |
| fl"fkrl difiac Annov IPvlnn hlrlanl*r Au .~ r4 (~i*~
| |
| . L* -- Ann~v ~st Fvalnn
| |
| ~ . ~. . *Muiri~ar U - - *~~*
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition:
| |
| Inability to control a key safety function from outside the Control Room.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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
| |
| " QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| " QOA 0010-05, Plant Operation with the Control Room Inaccessible AND
| |
| : 2. Control of ANY Table H1 key safety function is not reestablished in < 30 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.
| |
| Month 20XX QC 3-143 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)
| |
| Basis (cont):
| |
| 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 30 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-144 EP-AA-1006 (Revision XX)
| |
| | |
| FxA~lnn NnwIArler f3uad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:
| |
| * QCARP 0050-01, SB-i-i Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| * QOA 0010-05, Plant Operation with the Control Room Inaccessible 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-i-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-145 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Exelon Nuclear Nuclear Quad Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition:
| |
| FIRE potentially degrading the level of safety of the plant.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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)
| |
| " Main Control Room Envelope
| |
| * Unit and Shared Emergency Diesel Generator Rooms
| |
| * 4KV Switchgear Area
| |
| " Battery Rooms
| |
| " RHR Service Water Vaults
| |
| " Turbine Building Cable Tunnel
| |
| " Cribhouse 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 Month 20XX QC 3-146 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn Fxelnn Nuclear Nucleazr Ouadr Cities Annex RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-minutes of the initial report, alarm or indication.
| |
| OR
| |
| : 4. A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
| |
| 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.
| |
| Month 20XX QC 3-147 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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.
| |
| 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."
| |
| Month 20XX QC 3-148 EP-AA-1006 (Revision XX)
| |
| | |
| rl.r ar4(ifac~ Annoyv IFYalnn Niir~lanr
| |
| - Fv~Inn
| |
| ~ M.us-102r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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 HA3.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU4
| |
| : 2. UFSAR Section 3.2 Month 20XX QC 3-149 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Nuclear QuadCites AnexExelon RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition:
| |
| Seismic event greater than OBE levels.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| : 1. GGe~trel Roomn personnel fool an actual or potential sE;ismcce AND
| |
| : 2. The occUrrence Of a- eVent is confirmed in a Gesi manner deemed appropriate by the Shift Manager or Emergency Director.
| |
| Basis:
| |
| This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)1 . 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, 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 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.
| |
| Month 20XX QC 3-150 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY 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 Basis This 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 EAL wording 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.
| |
| Month 20XX QC 3-151 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU2
| |
| : 2. QCOA 0010-09, Earthquake Month 20XX QC 3-152 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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.
| |
| Operating Mode Applicability:
| |
| 4-1*2, 3, 4, &ý-D 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.
| |
| TableH3 Areas with Entry Related Mode Applicability A~ea Entry Related Mode
| |
| _ _ _ApplabMDy
| |
| _ _
| |
| ReacEto Bu ilding* Mode 3 and 4 Tue Bu*d~ng* _ _ _ _3
| |
| *Areas required to establish shutdown coo)inRg Table H3 Areas with Entry Related Mode Applicability Area Unit Entry Related Mode Applicability Reactor Building
| |
| * First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 Month 20XX QC 3-153 EP-AA-1006 (Revision XX)
| |
| | |
| fl"ari Mae Annov gI Ir% K-I-F n -&
| |
| hI- r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY AND
| |
| : 2. Entry into the room or area is prohibited or impeded 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).
| |
| 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.
| |
| Month 20XX QC 3-154 EP-AA-1 006 (Revision XX)
| |
| | |
| tliinrl Mina Annov I::Ynlen Kiil,-lar
| |
| ~- * .~
| |
| A ~'4-r;*;~a La~
| |
| * t* Annoy 0 ''* * *
| |
| * fl Fvalrhn. .Mm.r~Io~r
| |
| ~ . ~
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY
| |
| " 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 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 ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA5
| |
| : 2. UFSAR Section 3.2
| |
| : 3. ACIT 660892-20, Station Halon Discharge IDLH Evaluation Month 20XX QC 3-155 EP-AA-1006 (Revision XX)
| |
| | |
| IEx~lon NuclIazr Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 Initiating Condition:
| |
| Hazardous Event Operating Mode Applicability:
| |
| 1,2,3,4,5, 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 River level, as indicated by EITHER:
| |
| : a. High river water level > 594 ft.
| |
| OR
| |
| : b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam # 14 has failed.
| |
| Basis:
| |
| PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| Month 20XX QC 3-156 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
| |
| Month 20XX QC 3-157 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| Basis (cont):
| |
| 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 of power to a SAFETY SYSTEM comrponent as a result of leakage is an event of lesser impact and would be expected to cause small and localized damage. The consequence af thk tuine of event i6 adequaiitel" asese and ird~ressced fin Annn~r'-1nno- With' Technical Specifications.
| |
| 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, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
| |
| This EAL is not intended 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.
| |
| Month 20XX QC 3-158 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| Basis (cont):
| |
| EAL#5 Basis:
| |
| The Design Flood elevation is 594.5 ft. el. (rounded down to 594 ft. el. MSL). This initial design flood elevation is equal to the plant grade of 594.5 ft. el. and any mode of operation is, therefore, possible without additional protective measures.The station design is such that if Lock and Dam No. 14 were to fail, the water level would recede in the intake bay to the point where it would be separated from the river. As the water level recedes in the intake bay, circulating water, service water and fire diesel pumps would become inoperable, leaving only RHRSW and DGCW available to shutdown the units.
| |
| Use of the ultimate heat sink to shutdown the reactors requires the operation of portable diesel pumps with a total capacity of 5100 gpm to reverse the normal flow of makeup water. Makeup water would be provided from the river through the discharge piping and return to the river across the log boom in the intake bay.
| |
| Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, H or C.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU3
| |
| : 2. UFSAR Section 3.2
| |
| : 3. QCTP 0130-11 Internal Flood Protection Program
| |
| : 4. Drawing FL-1 Flood Barriers
| |
| : 5. Quad Cities Nuclear Power Station Unit 1 and 2 Internal Flooding Analysis Note Book, July 1993 Final Draft, prepared by Individual Plant Evaluation Partnership (IPEP)
| |
| Month 20XX QC 3-159 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Nuclear Quad Cities Annex Eeo ula 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, 4,5, 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 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 Month 20XX QC 3-160 EP-AA-1006 (Revision XX)
| |
| | |
| rliinfl 9'ifiaa Annoy I:=v,'lI*n M,,r-la-nr flu .~ r4 ~ Anngw ~vaJr~n kim ~
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Initiating Condition:
| |
| 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, 5, 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.
| |
| 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 a Site Area Emergency.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS7 Month 20XX QC 3-161 EP-AA-1006 (Revision XX)
| |
| | |
| Fxplnn NuJe.l*r Quad Cities Annex FYAIon Nuclerkr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Initiating Condition:
| |
| Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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.
| |
| 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 Month 20XX QC 3-162 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Applicability:
| |
| 1,2,3,4,5, 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:
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU7 Month 20XX QC 3-163 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:
| |
| * > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR 0 > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts
| |
| .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 ISFSls are covered under ICs HU1 and HAl.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 E-HU1
| |
| : 2. Certificate of Compliance No. 1014 Appendix A Section 5.7 Month 20XX QC 3-164 EP-AA-1006 (Revision XX)
| |
| | |
| l=xpInn NnJP.le_*r Duad Cities Annex Exelnn Nucler~I RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 Initiating Condition:
| |
| Release of gaseous radioactivity resulting in offsite dose greater than 1000 mRem TEDE or 5000 mRem thyroid CDE.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+09 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond 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.
| |
| Month 20XX QC 3-33 EP-AA-1006 (Revision XX)
| |
| | |
| FxA~lnn Nuclear Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 (cont)
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG1
| |
| : 2. EP-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 4. QGA 400 Radioactivity Release Control Month 20XX QC 3-34 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| * If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+08 uCi/sec for
| |
| > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| 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.
| |
| Month 20XX QC 3-35 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex QuadCites AnexExellon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RSl (cont)
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AS1
| |
| : 2. EP-AA-1 12-500 Emergency Environmental Monitoring
| |
| : 3. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-36 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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 Applicability:
| |
| 1,2, 3,4, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| " The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release 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. The sum of readings on the Rx Bldg and Chimney SPINGs > 3.84 E+07 uCi/sec for > 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
| |
| : a. > 10 mRem TEDE OR
| |
| : b. > 50 mRem CDE Thyroid OR Month 20XX QC 3-37 EP-AA-1006 (Revision XX)
| |
| | |
| Explon Eeo Nuclearula Quad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
| |
| : a. 10 mRem TEDE for 60 minutes of exposure OR
| |
| : b. 50 mRem CDE Thyroid for 60 minutes of exposure OR
| |
| : 4. Field survey results at or beyond the site boundary indicate EITHER:
| |
| : a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for > 60 minutes.
| |
| OR
| |
| : b. Analyses of field survey samples indicate > 50 mRem CDE Thyroid for 60 minutes of inhalation.
| |
| Basis:
| |
| This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of 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.
| |
| Escalation of the emergency classification level would be via IC RS1.
| |
| Month 20XX QC 3-38 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelnn Nuclear Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RAI (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA1
| |
| : 2. CY-QC-120-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. QOP 2000-24, Discharging to the River from the River Discharge Tank using the Waste Surge Pump
| |
| : 6. QOP 2000-25, Discharging to the River from the River Discharge Tank using the River Discharge Pump
| |
| : 7. CY-QC-120-729, Liquid Effluent Alarm Setpoints
| |
| : 8. QCOA 1700-02, High Radiation detected on Eberline Radiation Monitoring System
| |
| : 9. QCAN 912-5 C-6, Radwaste High Rad.
| |
| : 10. QCAN 901(2)-3-G-1, Liquid Process Rad. Monitor High Radiation
| |
| : 11. CY-QC-120-735, Main Chimney & Reactor Vent Noble Gas Release Rate Action Levels
| |
| : 12. QCOA 1700-01, Abnormal Chimney Radiation
| |
| : 13. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station
| |
| : 14. QGA 400 Radioactivity Release Control Month 20XX QC 3-39 EP-AA-1006 (Revision XX)
| |
| | |
| n"nd rifica Annoy PI=:,len hN"nlar Amm2rI
| |
| -- - - Ci*ia~
| |
| - * ** -- Ann~v g-* ** Fvolnn
| |
| ~st~ *
| |
| * U - ieIa~r
| |
| *
| |
| * NI. * - L4*
| |
| RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 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, 5, D Emergency Action Level (EAL):
| |
| Notes:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| " If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
| |
| " Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
| |
| : 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for > 60 minutes.
| |
| " Radwaste Effluent Monitor 1/2-1799-01 OR
| |
| " Discharge Permit specified monitor OR
| |
| : 2. The sum of readings on the Rx Bldg and Chimney SPINGs > 4.38 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).
| |
| OR
| |
| : 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
| |
| Basis:
| |
| This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
| |
| Month 20XX QC 3-40 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Ou~dCites AnexExellon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| EAL #1 Basis This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
| |
| The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.
| |
| EAL #2 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous effluent pathways.
| |
| EAL #3 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU1
| |
| : 2. CY-QC-1 20-729 Liquid Effluent Monitor Alarm Setpoints
| |
| : 3. CY-QC-120-737 Radioactive Liquid Discharge Batch Analysis
| |
| : 4. CY-QC-1 10-602 Radwaste System Sampling
| |
| : 5. EP-EAL-0606 Revision 2, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values Quad Cities Station Month 20XX QC 3-41 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, 6, 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.
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AG2 Month 20XX QC 3-42 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS2 Initiating Condition:
| |
| Spent fuel pool level at (site-specific Level 3 description).
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 6, D Emergency Action Level (EAL):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, AS2 Month 20XX QC 3-43 EP-AA-1006 (Revision XX)
| |
| | |
| I=xelon Nuclear Ouadr Cities Annex Exelon Nuclea~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS i RADIOLOGICAL EFFLUENTS RA2 Initiating Condition:
| |
| Significant lowering of water level above, or damage to, irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1,2, 3,4, 5, D Emergency Action Level (EAL):
| |
| : 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.
| |
| OR
| |
| : 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R1 Radiation Monitor reading >1000 mRem/hr.
| |
| OR
| |
| : 3. Lowering of spent fuel pool level to (site specific Level 2 value).
| |
| Table R1 Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Monitor
| |
| * 1(2) 1705-16B Fuel Pool Rad Monitor Basis:
| |
| REFUELING PATHWAY: all the cavities, tubes, canals and pools through which irradiated fuel may be moved or stored, but not including the reactor vessel below the flange.
| |
| IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
| |
| CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-44 EP-AA-1006 (Revision XX)
| |
| | |
| Oiiad Cities~ Annex Exellnn Nuclear I=x*lnn Nuclea~r RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 (cont)
| |
| Basis (cont):
| |
| 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 an 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 Ror C ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AA2
| |
| : 2. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity
| |
| : 3. QCAN 901(2)-3 B-1 Refuel Floor Hi Radiation
| |
| : 4. QCAN 901(2)-3 G-16/H-16 Fuel Pool Channel A/B Hi Radiation
| |
| : 5. QCIS 1700-07 Reactor Building Ventilation and Fuel Pool Radiation Monitoring Calibration and Functional Test Month 20XX QC 3-45 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 Initiating Condition:
| |
| UNPLANNED loss of water level above irradiated fuel.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| : 1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
| |
| * Refueling Cavity water level < 282 in. (Upper Wide range simulated signal).
| |
| OR
| |
| * Spent Fuel Pool water level < 19 ft. above the fuel (< - 4 ft. indicated level).
| |
| OR
| |
| * Indication or report of a drop in water level in the REFUELING PATHWAY.
| |
| AND
| |
| : b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table RI.
| |
| Table RI Fuel Handling Incident Radiation Monitors
| |
| * 1(2) 1705-16A Fuel Pool Rad Mon
| |
| * 1(2) 1705-16B Fuel Pool Rad Mon Month 20XX QC 3-46 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 (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.
| |
| 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 an 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, AU2
| |
| : 2. QCOP 0201-13 Reactor Level Upper Wide Range Reference Leg Extension Use and Control
| |
| : 3. Technical Specifications 3.7.8 Spent Fuel Storage Pool Water Level
| |
| : 4. Technical Specifications 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel
| |
| : 5. QCAN 901(2)-4 B-24 FUEL POOL STORAGE HI/LO LEVEL
| |
| : 6. QCOA 1900-01 Loss of Water Level in the Fuel Storage Pool or Reactor Cavity Month 20XX QC3-47 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS 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, 5, 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 R2 areas:
| |
| Table R2 Areas Requiring Continuous Occupancy
| |
| * Main Control Room (Unit 1 ARM Station #22)
| |
| * Central Alarm Station - (by survey)
| |
| OR
| |
| : 2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:
| |
| Table R3 Areas with Entry Related Mode Applicability Area Unit Entry Related Mode Applicability Reactor Building
| |
| " First Floor North Wall 1
| |
| " Second Floor North Wall 1 Mode 3 and 4
| |
| " First Floor South Wall 2
| |
| * Second Floor South Wall 2 High Pressure Heater Bay 1 &2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 Month 20XX QC 3-48 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nucilear RECOGNITION CATEGORY ABNORMAL RAD LEVELS I RADIOLOGICAL EFFLUENTS 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 R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.
| |
| This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
| |
| Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.
| |
| For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in 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.
| |
| * 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.).
| |
| Month 20XX QC 3-49 EP-AA-1006 (Revision XX)
| |
| | |
| Exnlon Exellon Nuclnnr Nuclenar Quad Cities Annex RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA3 (cont)
| |
| Basis (cont):
| |
| " 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. QCOP 1800-1 Operation of ARM Indicator/Trip Units
| |
| : 3. UFSAR Section 3.2
| |
| : 4. General Arrangement Drawings M-5, 6, 8 and 10 Month 20XX QC 3-50 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU3 Initiating Condition:
| |
| Reactor coolant activity greater than Technical Specification allowable limits.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| : 1. Offgas system radiation monitor HI-HI alarm.
| |
| OR
| |
| : 2. Specific coolant activity > 4.0 uCi/gm 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU3
| |
| : 2. Technical Specifications 3.4.6
| |
| : 3. Technical Specifications 3.7.6
| |
| : 4. QCAN 901(2)-3 C-2 OFF GAS HIGH-HIGH RADIATION Month 20XX QC 3-51 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:
| |
| Loss of ANY Two Barriers AND Loss or Potential Loss of the third barrier.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the General Emergency classification level each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-52 EP-AA-1006 (Revision XX)
| |
| | |
| 11iind Mfip4m Annoy FIylnn mnmirlar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:
| |
| Loss or Potential Loss of ANY two barriers.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
| |
| Basis:
| |
| Fuel Cladding, RCS and Containment comprise the fission product barriers.
| |
| At the Site Area Emergency classification level, each barrier is weighted equally.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-53 EP-AA-1 006 (Revision XX)
| |
| | |
| (3"nri Mtiac. AnnpX Pvalnn khirlanar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:
| |
| ANY Loss or ANY Potential Loss of EITHER Fuel Clad or RCS.
| |
| Operating Mode Applicability:
| |
| 1,2,3 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 Month 20XX QC 3-54 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Nuclear Quad Cities Annex Exelon NuclIear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:
| |
| RCS Activity Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Coolant activity > 300 uCi/gm Dose Equivalent 1-131.
| |
| Basis:
| |
| This threshold indicates that RCS radioactivity concentration is greater than 300 PCi/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
| |
| : 2. NF-AA-430, Failed Fuel Action Plan Month 20XX QC 3-55 EP-AA-1 006 (Revision XX)
| |
| | |
| tlimri riflac Annoy =,v,-lInn I*1" t-I--r flu .~ ii -...--
| |
| ---- Citia~ m.
| |
| Anngw ~vjt~iIw~n
| |
| ~ Mi- ar~I~r I ~S I I EU - l~~I RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:
| |
| RPV Water Level 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 > -142 inches (TAF)
| |
| OR
| |
| : 3. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.
| |
| This is identified in the BWROG EOPs/SAMGs 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 Month 20XX QC 3-56 EP-AA-1 006 (Revision XX)
| |
| | |
| Oiiad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)
| |
| Basis (cont):
| |
| RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
| |
| The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-57 EP-AA-1006 (Revision XX)
| |
| | |
| Explon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS Drywell radiation monitor reading > 6.65 E+02 R/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 300 [tCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
| |
| The radiation monitor reading in this threshold is higher than that specified for RCS Barrier RC5 Loss 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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. Core Damage Assessment Methodology (CDAM)
| |
| Month 20XX QC 3-58 EP-AA-1006 (Revision XX)
| |
| | |
| tl"nrl ifine Annoy Pvalnn Pd"d-lanr A. .~rI Citiac Ann~v ~vaInn Mmme-Ia2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-59 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. RPV water level cannot be restored and maintained > -142 inches (TAF)
| |
| OR
| |
| : 2. RPV water level cannot be determined.
| |
| Basis:
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.
| |
| The RPV water level threshold is the same as Fuel Clad 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 Month 20XX QC 3-60 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)
| |
| Basis (cont):
| |
| fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
| |
| Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.
| |
| In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level 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. QGA 100 RPV Control
| |
| : 3. QGA 500-4 RPV Flooding Month 20XX QC 3-61 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:
| |
| Primary Containment Pressure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell pressure >2.5 psig.
| |
| AND
| |
| : 2. Drywell pressure rise is due to RCS leakage.
| |
| Basis:
| |
| The > 2.5 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.
| |
| The second threshold focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Pressures of this magnitude can be caused by non-LOCA events such as a loss of Drywell cooling or inability to control primary containment vent/purge.
| |
| The release of mass from the RCS due to the as-designed/expected operation of any relief valve does not warrant an emergency classification.
| |
| A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| There is no Potential Loss threshold associated with Primary Containment Pressure.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 200 Primary Containment Control Month 20XX QC 3-62 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:
| |
| RCS Leak Rate Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. UNISOLABLE Main Steam Line (MSL), HPCI, Feedwater, RWCU or RCIC 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 > QGA 300 Maximum Normal operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300 Maximum Normal 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 on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break.
| |
| 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.
| |
| Month 20XX QC 3-63 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis (cont):
| |
| 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 safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the Torus, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.
| |
| Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.
| |
| A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
| |
| The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.
| |
| In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area. radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.
| |
| An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold
| |
| #1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
| |
| Month 20XX QC 3-64 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex OuadCitesRECOGNITION AnexExellon CATEGORY Nuclear FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. M-13 Main steam piping
| |
| : 3. UFSAR 5.2.5
| |
| : 4. QCOA 0201-01 Rev 016, Increasing Drywell Pressure
| |
| : 5. QOA 900-4 A-17 900-4 A-17 Annunciator
| |
| : 6. QCOS 1600-07 Reactor Coolant Leakage In The Drywell
| |
| : 7. QGA 300 Secondary Containment Control Month 20XX QC 3-65 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:
| |
| Primary Containment radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 1. Drywell radiation monitor reading > 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 Potential Loss threshold associated with Primary Containment Radiation.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2
| |
| : 2. CaIc. EP-EAL-0611 Month 20XX QC 3-66 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:
| |
| 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.
| |
| 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 Month 20XX QC 3-67 EP-AA-1006 (Revision XX)
| |
| | |
| Fxolnn N Hc.lo*r Quad Cities Annex Fxelnn Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:
| |
| RPV Water Level Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS 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. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 500-4 RPV Flooding Month 20XX QC 3-68 EP-AA-1006 (Revision XX)
| |
| | |
| i=x*lnn NHcle.nr Oun~d Cities Annex Fvnlnn NuclerIir RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:
| |
| Primary Containment 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 > 56 psig and rising.
| |
| OR
| |
| : 4. a. Drywell or torus hydrogen concentration > 6%.
| |
| AND
| |
| : b. Drywell or torus oxygen concentration > 5%.
| |
| OR
| |
| : 5. Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| Loss Threshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
| |
| These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.
| |
| Month 20XX QC 3-69 EP-AA-1006 (Revision XX)
| |
| | |
| f-)"nd rifiac Annov I=Y,-I~n I~el ln-r A *1214 ~ Anntw ~v~Ir~n MI.ria2r RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)
| |
| Basis (cont):
| |
| Potential Loss Threshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.
| |
| Potential Loss 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. UFSAR Fig. 6.2-16a
| |
| : 3. UFSAR Section 15.6
| |
| : 4. UFSAR 6.2.1.1
| |
| : 5. QGA 200, Primary Containment Control
| |
| : 6. Quad Cities PSTG Section 5, Primary Containment Control Month 20XX QC 3-70 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:
| |
| Primary Containment Radiation Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| POTENTIAL LOSS
| |
| : 1. Drywell radiation monitor reading > 1.55 E+03 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)
| |
| Month 20XX QC 3-71 EP-AA-1006 (Revision XX)
| |
| | |
| A. iiinf r~ifin Annav P~vlf~n Niu,,nlar RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:
| |
| Primary Containment Isolation Failure Operating Mode Applicability:
| |
| 1,2,3 Fission Product Barrier (FPB) Threshold:
| |
| LOSS
| |
| : 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 > QGA 300, Maximum Safe operating levels.
| |
| OR
| |
| : b. Secondary Containment area radiation level > QGA 300, Maximum Safe operating levels.
| |
| 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, HPCI or RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.
| |
| Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.
| |
| Month 20XX QC 3-72 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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 Month 20XX QC 3-73 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)
| |
| Basis (cont):
| |
| Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.
| |
| In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.
| |
| 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. QGA 200 Primary Containment Control
| |
| : 3. QGA 200-5 Hydrogen Control
| |
| : 4. QCOP 1600-13 Post-Accident Venting of the Primary Containment
| |
| : 5. QGA 300 Secondary Containment Control Month 20XX QC 3-74 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:
| |
| 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.
| |
| Basis:
| |
| 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, Table 9-F-2 Month 20XX QC 3-75 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex OundCites AnexExelon Nuclear 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. EITHER of the following:
| |
| : a. Restoration of at least one unit ECCS bus in < 1 hour is not likely.
| |
| OR
| |
| : b. RPV water level cannot be restored and maintained > -166 inches.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the SAMG program.
| |
| This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat 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 FGI. This will allow additional time for implementation of offsite protective actions.
| |
| Month 20XX QC 3-76 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)
| |
| Basis (cont):
| |
| Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
| |
| The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
| |
| The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. QGA 100 RPV Control Month 20XX QC 3-77 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lnn Nu*le_*r Quad Cities Annex Exelnn Nuclea~r 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:
| |
| 1,2,3 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| Basis:
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECOS. 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 or MG1.
| |
| Month 20XX QC 3-78 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lnn NH*lpar Ouadc Cities Annex Exelon Nucler~I RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-79 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit Auxiliary Transformer TR-11 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers AND
| |
| : 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.
| |
| 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 MUl.
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| * A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
| |
| Month 20XX QC 3-80 EP-AA-1006 (Revision XX)
| |
| | |
| Fxplnn NHc.le_*r ODiad C~ities Annex FvAlnn Nucleasr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)
| |
| Basis (cont):
| |
| * A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA1
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-81 EP-AA-1006 (Revision XX)
| |
| | |
| Oijad Cities Annex ExeIon Nuclear I=x*lon Nuc~lear 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:
| |
| 1,2,3 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 unit ECCS busses 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. (e.g. unit cross-tie breakers)
| |
| Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
| |
| Escalation of the emergency classification level would be via IC MAI.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SUW
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout Month 20XX QC 3-82 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex RECOGNITION CATEGORY Exelon Nuclear SYSTEM MALFUNCTIONS 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 unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO EDG 1(2) emergency diesel generators to supply power to vital busses.
| |
| AND
| |
| : 3. Voltage is < 105 VDC on 125 VDC battery busses #1 and #2.
| |
| AND
| |
| : 4. ALL AC and Vital DC power sources have been lost for >15 minutes.
| |
| 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 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 EALs are met.
| |
| Month 20XX QC 3-83 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lnn NHc.le_*r Quad Cities Annex Exellnn Nulairhr RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SG8
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| : 9. Technical Specifications 3.8.4 and B3.8.4
| |
| : 10. UFSAR Section 8.3.2
| |
| : 11. QOP 6900-02 125 VDC Electrical System
| |
| : 12. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-84 EP-AA-1006 (Revision XX)
| |
| | |
| I=xAInn N=m.lp_*r Oiiad Cities Annex FYAIon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2 Initiating Condition:
| |
| Loss of all vital DC power for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
| |
| Voltage is < 105 VDC on 125 VDC battery busses #1 and #2 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS8
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-85 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 > 5%.
| |
| AND
| |
| : 2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
| |
| AND
| |
| : 3. EITHER of the following conditions exist:
| |
| * RPV water level cannot be restored and maintained > -166 inches.
| |
| OR
| |
| * Heat Capacity Limit (QGA 200, Figure M) exceeded.
| |
| 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.
| |
| RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions.
| |
| Escalation of the emergency classification level would be via IC RG1 or FGI.
| |
| Month 20XX QC 3-86 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SS5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. QGA 200 Primary Containment Control Month 20XX QC 3-87 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2 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 > 5%.
| |
| AND
| |
| : 2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 5%.
| |
| Basis:
| |
| This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.
| |
| Month 20XX QC 3-88 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SA5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| Month 20XX QC 3-89 EP-AA-1 006 (Revision XX)
| |
| | |
| Duad Cities Annex Exellon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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
| |
| > 5%.
| |
| 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
| |
| > 5%.
| |
| 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 control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
| |
| Month 20XX QC 3-90 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram / ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram / ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
| |
| A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
| |
| Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.
| |
| The plant response to the failure of an automatic or manual reactor 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 control 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 FA1. 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.
| |
| Month 20XX QC 3-91 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities Annex Exallon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, SU5
| |
| : 2. QGA 100 RPV Control
| |
| : 3. QGA 101 RPV Control (ATWS)
| |
| : 4. Technical Specifications Table 3.3.1.1-1
| |
| : 5. Technical Specification 3.3.1.3
| |
| : 6. Technical Specification Bases 3.3.1.1 and 3.3.1.3 Month 20XX QC 3-92 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| * Reactor Power
| |
| * RPV Water Level
| |
| * RPV Pressure
| |
| * Drywell Pressure
| |
| " Torus Level
| |
| " Torus Temperature AND
| |
| : 2. ANY Table M2 transient in progress.
| |
| Table M2 Significant Transients
| |
| " Turbine Trip
| |
| * Reactor Scram
| |
| * ECCS Activation
| |
| * Recirc. Runback > 25% Reactor Power Change
| |
| " Thermal Power oscillations > 10% Reactor Power Change Month 20XX QC 3-93 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear 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 EGGS. 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 Month 20XX QC 3-94 EP-AA-1006 (Revision XX)
| |
| | |
| f_)"nri rifina Annov I=Yalnn N,,r~lanr A.
| |
| - - u~
| |
| - -ri (~iti~
| |
| - . S. -- ~Annoy U*
| |
| * S FvciInn
| |
| - MImr102r U - ~*
| |
| RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 parameter from within the Control Room for > 15 minutes.
| |
| Table M1 Control Room Parameters
| |
| * Reactor Power
| |
| * RPV Water Level
| |
| * RPV Pressure
| |
| * Drywell Pressure
| |
| * Torus Level
| |
| * Torus Temperature Basis:
| |
| UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.
| |
| As used in this EAL, an "inability to monitor" means that values for 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.
| |
| Month 20XX QC 3-95 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex I=x*lon Nucler~I Nuclear Ouad itiesAnnexExeIon 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 Month 20XX QC 3-96 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 Initiating Condition:
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-97 EP-AA-1006 (Revision XX)
| |
| | |
| I=x*lon NucIp_*r Ouaid C~ities~Annex ExeInn Nucrlear 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.
| |
| This EAL #2a 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.
| |
| This 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 is not 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 SA9 Month 20XX QC 3-98 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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.
| |
| 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 Month 20XX QC 3-99 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)
| |
| Basis (cont):
| |
| 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 Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.
| |
| 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. QCOS 1600-07 Reactor Coolant Leakage in the Drywell
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QCOA 0201-01 Increasing Drywell Pressure Month 20XX QC 3-100 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelnn Nuclear I=x*lon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Initiating Condition:
| |
| Loss of all On-site or Off-site communications capabilities.
| |
| Operating Mode Applicability:
| |
| 1,2,3 Emergency 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 Radio X Plant Page X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite 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.).
| |
| Month 20XX QC 3-101 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis This EAL addresses a total loss of the communications methods used in support of routine plant operations.
| |
| EAL #2 Basis This EAL 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) Form.
| |
| EAL #3 Basis This EAL 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-102 EP-AA-1006 (Revision XX)
| |
| | |
| IFxAlnn NurlIa~r Duad Cities Annex Exelon Nuclear 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:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. Loss of ALL offsite AC power to unit ECCS busses.
| |
| AND
| |
| : 2. Failure of Unit EDG 1(2), shared EDG 1/2 and SBO DG 1(2) emergency diesel generators to supply power to unit ECCS busses.
| |
| AND
| |
| : 3. Failure to restore power to at least one unit ECCS bus in < 15 minutes from the time of loss of both offsite and onsite AC power.
| |
| 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.
| |
| Month 20XX QC 3-103 EP-AA-1006 (Revision XX)
| |
| | |
| Duad Cities. Annex OuadCites AnexExellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA2
| |
| : 2. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "ACTURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-104 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS cul Initiating Condition:
| |
| Loss of all but one AC power source to emergency buses for 15 minutes or longer.
| |
| Operating Mode Applicability:
| |
| 4,5, D Emergency Action Level (EAL):
| |
| Note:
| |
| * The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
| |
| : 1. AC power capability to unit ECCS busses reduced to only one of the following power sources for > 15 minutes.
| |
| * Reserve auxiliary Transformer TR-12 (TR-22)
| |
| * Unit auxiliary transformer TR-1 1 (TR-21)
| |
| * Unit Emergency Diesel Generator
| |
| * Shared Emergency Diesel Generator
| |
| * Station Blackout Diesel Generator
| |
| * Unit crosstie breakers 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.
| |
| Month 20XX QC 3-105 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU1 (cont)
| |
| Basis (cont):
| |
| An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
| |
| " 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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. UFSAR Figure 8.3-1
| |
| : 3. UFSAR Section 8.3
| |
| : 4. QCOA 6100-03 Loss of Offsite Power
| |
| : 5. QOP 6100-02 Restoring Reserve Auxiliary Transformer 12 To Service
| |
| : 6. QOP 6100-04 Restoring Reserve Auxiliary Transformer 22 To Service
| |
| : 7. QCOA 6100-04 Station Blackout
| |
| : 8. GE letter No. 92-38 from L.G. Knutson to Pat Donahue, dated April 7, 1992, "AC TURBINE LOADS SMALL TASK NO. QC107" (Station Blackout analysis)
| |
| Month 20XX QC 3-106 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:
| |
| Hazardous event affecting SAFETY SYSTEM required for the current operating mode.
| |
| Operating Mode Applicability:
| |
| 4,5 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.
| |
| 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.
| |
| Month 20XX QC 3-107 EP-AA-1006 (Revision XX)
| |
| | |
| A, .rl riCiiac Annoy Pvnlrn Khid-In!mr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (cont)
| |
| Basis (cont):
| |
| 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, 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA6 Month 20XX QC 3-108 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:
| |
| Loss of Vital DC power 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.
| |
| Voltage is < 105 VDC on required 125 VDC battery busses #1 and #2 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 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.
| |
| Month 20XX QC 3-109 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CU4
| |
| : 2. Technical Specifications 3.8.4 and B3.8.4
| |
| : 3. UFSAR Section 8.3.2
| |
| : 4. QOP 6900-02 125 VDC Electrical System
| |
| : 5. QCTS 0230-01 Unit One (Two) 125 VDC Service Test Normal or Alternate Battery Month 20XX QC 3-110 EP-AA-1006 (Revision XX)
| |
| | |
| Mind Mae Annov I,-Invlrn M, "Alr fl112r4
| |
| - - 4 (~i*iaQ
| |
| -
| |
| * t* S - Ann~v S~* U* *fl ~vaIr~n fl * * * *Mi U - ir~Ia~r
| |
| * 4*
| |
| RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:
| |
| Loss of all onsite or offsite communications capabilities.
| |
| Operating Mode Applicability:
| |
| 4,5, 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 Radio X Plant Page X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:
| |
| This IC addresses a significant loss of on-site or offsite communications capabilities.
| |
| While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Outside 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.
| |
| Month 20XX QC 3-111 EP-AA-1006 (Revision XX)
| |
| | |
| tl"nri Pifipc Annov I=Yaltnn NIuid-lar 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-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) 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. EP-MW-124-1001 Facilities Inventories and Equipment Tests
| |
| : 3. UFSAR Section 9.5.2 Month 20XX QC 3-112 EP-AA-1006 (Revision XX)
| |
| | |
| tliumrl ritioa Annoy F=volnn N"Ii-lanr RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 Initiating Condition:
| |
| Inability to maintain the plant in cold shutdown.
| |
| 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.
| |
| " 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*
| |
| Not Intact Established 20 minutes*
| |
| 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.
| |
| 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.
| |
| 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.).
| |
| Month 20XX QC 3-113 EP-AA-1006 (Revision XX)
| |
| | |
| Mind Mac Annov I:::a--In N"Aal--r A m.2,4 Ci$i~ Annav Fv~Inn NiirIa2r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| Month 20XX QC 3-114 EP-AA-1006 (Revision XX)
| |
| | |
| Otiad Cities Annex Exellon Nuclear I=x*lon Nucla~r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CA3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. OU-AA-103 Shutdown Safety
| |
| : 6. QCOA 1000-02 Loss of Shutdown Cooling
| |
| : 7. QGA 100 RPV Control
| |
| : 8. QGA 100 RPV Control Detail A
| |
| : 9. QCGP 1-1 Normal Unit Startup
| |
| : 10. QCIS 0600-01 Unit One Division 1 Reactor Pressure 0 to 1200 psig Indication Calibration Month 20XX QC 3-115 EP-AA-1006 (Revision XX)
| |
| | |
| I:x*lnn Nuclear Oiiad Cities Annex Exelnn Nuclea~r RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 Initiating Condition:
| |
| UNPLANNED rise in RCS temperature 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.
| |
| * 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.).
| |
| A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
| |
| Month 20XX QC 3-116 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 (cont)
| |
| Basis (cont):
| |
| 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):
| |
| : 1. NEI 99-01 Rev 6, CU3
| |
| : 2. Technical Specifications Table 1.1-1
| |
| : 3. QGA 100, RPV Control
| |
| : 4. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 5. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 6. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge
| |
| : 7. QCOA 1000-02 Loss of Shutdown Cooling Month 20XX QC 3-117 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Initiating Condition:
| |
| Loss of RPV inventory affecting fuel clad integrity with containment challenged.
| |
| 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 a. RPV water level < -142 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
| |
| " Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-118 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| Table C4 Containment Challenge Indications
| |
| " Primary Containment Hydrogen Concentration > 6% and Oxygen > 5%
| |
| * UNPLANNED rise in containment pressure
| |
| * CONTAINMENT CLOSURE not established*
| |
| * ANY Secondary Containment radiation monitor > QGA 300, Maximum Safe operating level.
| |
| * if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.
| |
| 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.
| |
| Month 20XX QC 3-119 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis (cont):
| |
| 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-120 EP-AA-1006 (Revision XX)
| |
| | |
| l:y*Inn N.*lp_;*r Ouasd Cities Annex FvAllnn Nula~r~I RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CG1
| |
| : 2. QGA 100, RPV Control
| |
| : 3. Technical Specifications 3.3.1
| |
| : 4. Technical Specifications 3.6.1.1
| |
| : 5. Technical Specifications 3.6.4.1
| |
| : 6. QGA-200-5, Hydrogen Control
| |
| : 7. UFSAR 6.2.1.1
| |
| : 8. QGA 300 Secondary Containment Control
| |
| : 9. EP-EAL-0501, Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Month 20XX QC 3-121 EP-AA-1006 (Revision XX)
| |
| | |
| Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 Initiating Condition:
| |
| Loss of RPV inventory affecting core decay heat removal capability.
| |
| 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. With CONTAINMENT CLOSURE not established, RPV water level
| |
| <- 65 inches OR
| |
| : 2. With CONTAINMENT CLOSURE established, RPV water level < - 142 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
| |
| * Fuel Handling ARM 1(2)-1705-16A or B >3000 mR/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 in RPV inventory Month 20XX QC 3-122 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN I REFUELING SYSTEM MALFUNCTIONS CS6 (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.
| |
| 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.b and 2.b 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 tank levels. Sump and/or tank 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 RG1.
| |
| Month 20XX QC 3-123 EP-AA-1006 (Revision XX)
| |
| | |
| E~In~ln Exellon Nclea~r Nuclear Ouad ities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, CS1
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.6.1.1
| |
| : 4. Technical Specifications 3.6.4.1
| |
| : 5. QGA 100, RPV Control
| |
| : 6. Technical Specifications 3.3.1
| |
| : 7. Technical Specifications Table 3.3.3.1-1
| |
| : 8. Technical Specifications 3.3.5.1
| |
| : 9. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 10. Technical Specifications 3.4.4
| |
| : 11. UFSAR 5.2.5
| |
| : 12. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 13. QOA 900-4 A-17, Annuciator Response Month 20XX QC 3-124 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:
| |
| Loss of RPV 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 < - 59 inches.
| |
| OR
| |
| : 2. a. RPV water level cannot be monitored for > 15 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 in 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 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.
| |
| Month 20XX QC 3-125 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 (cont)
| |
| Basis (cont):
| |
| EAL #1 Basis A lowering of water level below -59 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 decrease in water level will lead to core uncovery.
| |
| 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). An 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 tank levels. Sump and/or tank 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 Specifications 3.3.5.1
| |
| : 3. QCOS 1600-07, Reactor Coolant Leakage in the Drywell
| |
| : 4. Technical Specifications 3.4.4
| |
| : 5. UFSAR 5.2.5
| |
| : 6. QCOA 0201-01, Increasing Drywell Pressure
| |
| : 7. QOA 900-4 A-17, Annuciator Response
| |
| : 8. QGA 100, RPV Control
| |
| : 9. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 10. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 11. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-126 EP-AA-1006 (Revision XX)
| |
| | |
| Ouad Cities Annex QuadCites AnexExelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 Initiating Condition:
| |
| UNPLANNED loss of RPV 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 water level to 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.
| |
| 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 in RPV inventory Month 20XX QC 3-127 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (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 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.
| |
| 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 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 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 tank levels. Sump and/or tank 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.
| |
| Month 20XX QC 3-128 EP-AA-1 006 (Revision XX)
| |
| | |
| Exallon ula Eeo Nuclear Quad Cities Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev. 6 CUI
| |
| : 2. Technical Specifications 3.3.5.1
| |
| : 3. Technical Specifications 3.4.4
| |
| : 4. UFSAR 5.2.5
| |
| : 5. QGA 100, RPV Control
| |
| : 6. QCOP 0201-02, Filling the Reactor Vessel and/or Reactor Cavity Using a Condensate Booster Pump via the Feedwater System
| |
| : 7. QCOP 0201-13, Reactor Vessel Upper Wide Range Reference Leg Extension Use and Control
| |
| : 8. QCOP 0201-14, Reactor Vessel Level Control Using a Local Pressure Gauge Month 20XX QC 3-129 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, 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 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)
| |
| * RPV Water Level (ability to cool the core)
| |
| * RCS Heat Removal (ability to maintain heat sink)
| |
| 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.
| |
| 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.
| |
| Month 20XX QC 3-130 EP-AA-1 006 (Revision XX)
| |
| | |
| t'J".qd riflac Anna-z I=v,',Inn M,,rl-lar RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 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 ContingencyPlan [and Independent Spent Fuel Storage Installation Security Program].
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev. 6 HG1
| |
| : 2. Station Security Plan - Appendix C Month 20XX QC 3-131 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:
| |
| HOSTILE ACTION within the PROTECTED AREA.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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].
| |
| Month 20XX QC 3-132 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 (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 Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HS1
| |
| : 3. Station Security Plan - Appendix C Month 20XX QC 3-133 EP-AA-1 006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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.
| |
| Month 20XX QC 3-134 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (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 QCOA 0010-20, Security Event.
| |
| EAL #2 Basis Is 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.
| |
| Month 20XX QC 3-135 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01 Rev 6, HA1
| |
| : 2. Station Security Plan - Appendix C
| |
| : 3. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-136 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:
| |
| Confirmed SECURITY CONDITION or threat.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| Basis:
| |
| 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.
| |
| Month 20XX QC 3-137 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HUl (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-101-132.
| |
| 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 QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| 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, HU1
| |
| : 2. SY-AA-1 01-132, Security Assessment and Response to Unusual Activities
| |
| : 3. Station Security Plan - Appendix C
| |
| : 4. NRC Safeguards Advisory 10/6/01
| |
| : 5. Letter from Mr. B. A. Boger (NRC) to Ms. Lynette Hendricks (NEI) dated 2/4/02
| |
| : 6. QCOA 0010-20, Security Event (G.7.A, G.7.B, G.7.C, G.7.D, G.7.K)
| |
| Month 20XX QC 3-138 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition:
| |
| Inability to control a key safety function from outside the Control Room.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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
| |
| * QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| * QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR QOA 0010-05, Plant Operation with the Control Room Inaccessible AND
| |
| : 2. Control of ANY Table H1 key safety function is not reestablished in < 30 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.
| |
| Month 20XX QC 3-139 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)
| |
| Basis (cont):
| |
| 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 30 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-i-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-i-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-140 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, D Emergency Action Level (EAL):
| |
| A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:
| |
| * QCARP 0050-01, SB-i-i Injection with SSMP and Bringing the Unit to Cold Shutdown OR
| |
| " QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown OR
| |
| " QOA 0010-05, Plant Operation with the Control Room Inaccessible 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA6
| |
| : 2. QOA 0010-05, Plant Operation with the Control Room Inaccessible
| |
| : 3. QCARP 0050-01, SB-I-1 Injection with SSMP and Bringing the Unit to Cold Shutdown
| |
| : 4. QCARP 0050-02, SB-I-1 Injection with RCIC and Bringing the Unit to Cold Shutdown Month 20XX QC 3-141 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition:
| |
| FIRE potentially degrading the level of safety of the plant.
| |
| Operating Mode Applicability:
| |
| 1,2, 3, 4, 5, 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)
| |
| " Main Control Room Envelope
| |
| " Unit and Shared Emergency Diesel Generator Rooms
| |
| * 4KV Switchgear Area
| |
| * Battery Rooms
| |
| " RHR Service Water Vaults
| |
| " Turbine Building Cable Tunnel
| |
| " Cribhouse OR
| |
| : 2. a. Receipt of a sinale 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 Month 20XX QC 3-142 EP-AA-1006 (Revision XX)
| |
| | |
| I=Y*lnn Nirl*r 0.32,1 ritip Annpw FvalInn N"A~canr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Emergency Action Level (EAL) (cont):
| |
| : 3. A FIRE within the plant or ISFSI PROTECTED AREA not extinguished in < 60-minutes of the initial report, alarm or indication.
| |
| OR
| |
| : 4. A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
| |
| 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.
| |
| Month 20XX QC 3-143 EP-AA-1 006 (Revision XX)
| |
| | |
| Otiad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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.
| |
| 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."
| |
| Month 20XX QC 3-144 EP-AA-1 006 (Revision XX)
| |
| | |
| ('1"nei Mae Annav Pvnlrn Kl"Anmr RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)
| |
| Basis (cont):
| |
| 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.
| |
| 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 HA3.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU4
| |
| : 2. UFSAR Section 3.2 Month 20XX QC 3-145 EP-AA-1006 (Revision XX)
| |
| | |
| (_)"nri rifine Annav Pvalnn kl"rlanr A I .2,4 Citia~ Ann~w Fvoinn MIIrI~h2r RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition:
| |
| Seismic event greater than OBE levels.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| Basis:
| |
| This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)1 . 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, 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.
| |
| 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 public 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.
| |
| Month 20XX QC 3-146 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon NucerIAm RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)
| |
| Basis (cont):
| |
| 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 EAL wording 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.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU2
| |
| : 2. QCOA 0010-09, Earthquake
| |
| : 3. US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions.
| |
| Month 20XX QC 3-147 EP-AA-1006 (Revision XX)
| |
| | |
| ir4 Min ifac Anncov PvI::lnn Nelalr 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.
| |
| 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 Unit Entry Related Mode Applicability Reactor Building
| |
| " First Floor North Wall 1
| |
| * Second Floor North Wall 1 Mode 3 and 4
| |
| * First Floor South Wall 2
| |
| " Second Floor South Wall 2 High Pressure Heater Bay 1&2 MSIV Room 1 Mode 3 Second Floor Turbine Bldg. N.E. Corner 2 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.
| |
| Month 20XX QC 3-148 EP-AA-1006 (Revision XX)
| |
| | |
| I=Y*_lnn NuJi*.lm*ar 0 Ilad C~ities Annex Fvazimnn NKwIAria RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)
| |
| Basis (cont):
| |
| 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 ifthis 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).
| |
| 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.
| |
| Month 20XX QC 3-149 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)
| |
| Basis (cont):
| |
| 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 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 ICs.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HA5
| |
| : 2. UFSAR Section 3.2
| |
| : 3. ACIT 660892-20, Station Halon Discharge IDLH Evaluation Month 20XX QC 3-150 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 Initiating Condition:
| |
| Hazardous Event Operating Mode Applicability:
| |
| 1,2,3,4,5, 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 River level, as indicated by EITHER:
| |
| : a. High river water level > 594 ft.
| |
| OR
| |
| : b. Report of substantial reduction in river level by site personnel and confirmation by the Corp. of Engineers that Dam # 14 has failed.
| |
| Basis:
| |
| PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.
| |
| SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.
| |
| Month 20XX QC 3-151 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exellon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| 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.
| |
| 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, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
| |
| This EAL is not intended 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.
| |
| Month 20XX QC 3-152 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)
| |
| Basis (cont):
| |
| EAL#5 Basis:
| |
| The Design Flood elevation is 594.5 ft. el. (rounded down to 594 ft. el. MSL). This initial design flood elevation is equal to the plant grade of 594.5 ft. el. and any mode of operation is, therefore, possible without additional protective measures.The station design is such that if Lock and Dam No. 14 were to fail, the water level would recede in the intake bay to the point where it would be separated from the river. As the water level recedes in the intake bay, circulating water, service water and fire diesel pumps would become inoperable, leaving only RHRSW and DGCW available to shutdown the units.
| |
| Use of the ultimate heat sink to shutdown the reactors requires the operation of portable diesel pumps with a total capacity of 5100 gpm to reverse the normal flow of makeup water. Makeup water would be provided from the river through the discharge piping and return to the river across the log boom in the intake bay.
| |
| Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, H or C.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HU3
| |
| : 2. UFSAR Section 3.2
| |
| : 3. QCTP 0130-11 Internal Flood Protection Program
| |
| : 4. Drawing FL-1 Flood Barriers
| |
| : 5. Quad Cities Nuclear Power Station Unit 1 and 2 Internal Flooding Analysis Note Book, July 1993 Final Draft, prepared by Individual Plant Evaluation Partnership (IPEP)
| |
| Month 20XX QC 3-153 EP-AA-1006 (Revision XX)
| |
| | |
| Quad Cities Annex Exelon Nuclear 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:
| |
| 1,2,3,4,5, 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 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 Month 20XX QC 3-154 EP-AA-1006 (Revision XX)
| |
| | |
| rl"nrl Mac Annov I=vnlrn kl"Al--r fl.m~mrI
| |
| ~- S.- Ci$i~~
| |
| W ** -- Annoy C-.. * * . ~efl ~voInn
| |
| ~ . ~. . MuurI~~r I. - ~
| |
| RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Initiating Condition:
| |
| Other conditions exist which in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY.
| |
| Operating Mode Applicability:
| |
| 1,2,3,4,5, 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.
| |
| 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 a Site Area Emergency.
| |
| Basis Reference(s):
| |
| : 1. NEI 99-01, Rev 6 HS7 Month 20XX QC 3-155 EP-AA-1006 (Revision XX)
| |
| | |
| Fx*_lnn N ilrulI_*r 011ad Nties Annex Fxelon Nnelpar RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Initiating Condition:
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| Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT Operating Mode Applicability:
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| 1,2,3,4,5, D Emergency Action Level (EAL):
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| 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.
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| Basis:
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| 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).
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| 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.
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| 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.
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| Basis Reference(s):
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| : 1. NEI 99-01, Rev 6 HA7 Month 20XX QC 3-156 EP-AA-1006 (Revision XX)
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| Quad Cities Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU7 Initiating Condition:
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| Other conditions exist which in the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.
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| Operating Mode Applicability:
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| 1,2, 3, 4, 5, D Emergency Action Level (EAL):
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| 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.
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| Basis:
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| 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.
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| Basis Reference(s):
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| : 1. NEI 99-01, Rev 6 HU7 Month 20XX QC 3-157 EP-AA-1006 (Revision XX)
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| I=x*_lnn NH*I*r (3.q~d Cities~Annex FYAlnn Nucler~I RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.
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| Operating Mode Applicability:
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| 1,2,3,4,5, D Emergency Action Level (EAL):
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| Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:
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| 0 > 40 mr/hr (gamma + neutron) on the top of the spent fuel cask OR 0 > 220 mr/hr (gamma + neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts
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| .Basis:
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| CONFINEMENT BOUNDARY: The irradiated fuel dry storage cask barrier(s) between areas containing radioactive substances and the environment.
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| 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.
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| 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.
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| 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.
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| Security-related events for ISFSs are covered under ICs HU1 and HAl.
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| Basis Reference(s):
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| : 1. NEI 99-01, Rev 6 E-HU1
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| : 2. Certificate of Compliance No. 1014 Appendix A Section 5.7 Month 20XX QC 3-158 EP-AA-1006 (Revision XX)}}
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