ML17164A148
| ML17164A148 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/31/2017 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17164A149 | List: |
| References | |
| NMP1L3158 | |
| Download: ML17164A148 (130) | |
Text
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-25 EP-AA-10XX (Rev X)
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, 6, 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monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.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. Reading on ANY Table R1 Effluent Monitor > Table R1 column GE value for
> 15 minutes.
- 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
- 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.
- b. Analyses of field survey samples indicate > 5000 mRem Thyroid CDE for 60 minutes of inhalation.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-26 EP-AA-10XX (Rev X)
RG1 (cont)
Emergency Action Level (EAL) (cont):
Table R1 - Effluent Monitor Thresholds Monitor General Emergency (GE)
CNMT Vent Noble Gas High Range (R-12A) 2.56 E +01 uCi/cc Plant Vent Noble Gas High Range (R-14A) 1.41 E +02 uCi/cc Air Ejector Noble Gas High Range (R-48) 2.17 E +03 uCi/cc Basis:
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA PAG of 1000 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.
R.E. Ginna Nuclear Power Plant Off-Site Dose Calculation Manual (ODCM)
- 3.
EP-EAL-0632, Calculation of Ginna Nuclear Power Station Table R-1 EAL Threshold Values
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-27 EP-AA-10XX (Rev X)
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, 6, 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. Reading on ANY Table R1 Effluent Monitor > Table R1 column SAE value for
> 15 minutes.
- 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
- 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.
- b. Analyses of field survey samples indicate > 500 mRem Thyroid CDE for 60 minutes of inhalation.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-28 EP-AA-10XX (Rev X)
RS1 (cont)
Emergency Action Level (EAL) (cont):
Table R1 - Effluent Monitor Thresholds Monitor Site Area Emergency (SAE)
CNMT Vent Noble Gas High Range (R-12A) 2.56 E +00 uCi/cc Plant Vent Noble Gas High Range (R-14A) 1.41 E +01 uCi/cc Air Ejector Noble Gas High Range (R-48) 2.17 E +02 uCi/cc Basis:
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 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.
R.E. Ginna Nuclear Power Plant Off-Site Dose Calculation Manual (ODCM)
- 3.
EP-EAL-0632, Calculation of Ginna Nuclear Power Station Table R-1 EAL Threshold Values
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-29 EP-AA-10XX (Rev X)
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, 6, 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. Reading on ANY Table R1 Effluent Monitor > Table R1 column Alert value for
> 15 minutes.
- 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
- a.
- b.
- 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
- b. 50 mRem Thyroid CDE for 60 minutes of exposure
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-30 EP-AA-10XX (Rev X)
RA1 (cont)
Emergency Action Level (EAL) (cont):
- 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.
- b. Analyses of field survey samples indicate > 50 mRem Thyroid CDE for 60 minutes of inhalation.
Table R1 - Effluent Monitor Thresholds Monitor Alert CNMT Vent Noble Gas High Range (R-12A) 2.56 E -01 uCi/cc Plant Vent Noble Gas High Range (R-14A) 1.41 E +00 uCi/cc Air Ejector Noble Gas High Range (R-48) 2.17 E +01 uCi/cc 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.
This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.
Escalation of the emergency classification level would be via IC RS1.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-31 EP-AA-10XX (Rev X)
RA1 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, AA1
- 2.
R.E. Ginna Nuclear Power Plant Off-Site Dose Calculation Manual (ODCM)
- 3.
EP-EAL-0632, Calculation of Ginna Nuclear Power Station Table R-1 EAL Threshold Values
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-32 EP-AA-10XX (Rev X)
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, 6, 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 > 2x the alarm set point established by a current radioactive release discharge permit for 60 minutes.
Liquid Radwaste Effluent Monitor (R-18) with no isolation Discharge Permit specified monitor OR
- 2. Reading on ANY Table R-1 effluent monitors > Table R-1 column UE value for 60 minutes.
- 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 x ODCM Limit with a release duration of > 60 minutes.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-33 EP-AA-10XX (Rev X)
RU1 (cont)
Emergency Action Level (EAL) (cont):
Table R1 - Effluent Monitor Thresholds Monitor Unusual Event (UE)
CNMT Vent Noble Gas High Range (R-12A)
N/A Plant Vent Noble Gas High Range (R-14A) 3.63 E -01 uCi/cc Air Ejector and Gland Steam exhaust Monitor (R-15) 2.29 E +06 cpm Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
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 for > 60 minutes. 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-34 EP-AA-10XX (Rev X)
RU1 (cont)
Basis (cont):
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.
R.E. Ginna Nuclear Power Plant Off-Site Dose Calculation Manual (ODCM)
- 3.
EP-EAL-0632, Calculation of Ginna Nuclear Power Station Table R-1 EAL Threshold Values
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-35 EP-AA-10XX (Rev X)
RG2 Initiating Condition:
Spent fuel pool level cannot be restored to at least 251.67 ft. 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 251.67 ft. as indicated on EITHER LI-310 or LI-311 for > 60 minutes.
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
- 2.
WPLNRC-1002747 Letter from E.D. Dean (CENG) to document control desk (NRC), six-month status report in response to March 12, 2012 commission order modifying license with regard to reliable Spent Fuel Pool Instrumentation (Order number EA-12-051) dated August 27, 2013 (ML13254A279)
- 3.
ECP-13-000547-015-7-01 SFP Instrumentation Design Change Technical Evaluation
- 4.
66-9218244 Factory Acceptance Test Report for SFP Level Instrumentation System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-36 EP-AA-10XX (Rev X)
RS2 Initiating Condition:
Spent fuel pool level at 251.67 ft.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
Lowering of spent fuel pool level to 251.67 ft. as indicated on EITHER LI-310 or LI-311.
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
- 2.
WPLNRC-1002747 Letter from E.D. Dean (CENG) to document control desk (NRC), six-month status report in response to March 12, 2012 commission order modifying license with regard to reliable Spent Fuel Pool Instrumentation (Order number EA-12-051) dated August 27, 2013 (ML13254A279)
- 3.
ECP-13-000547-015-7-01 SFP Instrumentation Design Change Technical Evaluation
- 4.
66-9218244 Factory Acceptance Test Report for SFP Level Instrumentation System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-37 EP-AA-10XX (Rev X)
RA2 Initiating Condition:
Significant lowering of water level above, or damage to, irradiated fuel.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
- 1. Uncovery of irradiated fuel in the REFUELING PATHWAY.
- 2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R2 Radiation Monitor Alarm.
- 3. Lowering of spent fuel pool level to 257.25 ft. as indicated on EITHER LI-310 or LI-311.
Table R2 Fuel Handling Radiation Monitors R-2 Containment (Mode 6 and D)
R-5 Spent Fuel Pool (All Modes)
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-38 EP-AA-10XX (Rev X)
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 a rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes.
EAL #2 Basis:
This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
EAL #3 Basis:
Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.
Escalation of the emergency would be based on either Recognition Category R or C ICs.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-39 EP-AA-10XX (Rev X)
RA2 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, AA2
- 2.
ER-SFP.1 Loss of SFP Cooling
- 3.
ER-SFP.2 Diverse SFP Makeup and Spray
- 4.
Technical Specifications Section 3.7.11 Spent Fuel Pool (SFP) Water Level
- 5.
Technical Specifications Section 3.9.6 Refueling Water Cavity Level
- 6.
P-9 Radiation Monitoring System
- 7.
FL-13-036 Letter from E.D. Dean (CENG) to document control desk (NRC), six-month status report in response to March 12, 2012 commission order modifying license with regard to reliable Spent Fuel Pool Instrumentation (Order number EA-12-051) dated August 27, 2013
- 8.
EPC-13-000547-015-7-01 SFP Instrumentation Design Change Technical Evaluation
- 9.
66-9218244 Factory Acceptance Test Report for SFP Level Instrumentation System
- 10.
CALC-2013-0001 Ginna Spent Fuel Area Dose Rates at Severely Reduced Water Levels
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-40 EP-AA-10XX (Rev X)
RU2 Initiating Condition:
UNPLANNED loss of water level above irradiated fuel.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, 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 < 23 ft. above the Reactor Flange.
OR Spent Fuel Pool water level < 23 ft. (Equivalent to < 274.50 feet on EITHER LI-310 or LI-311) above the fuel.
OR Indication or report of a drop in water level in the REFUELING PATHWAY.
AND
- b. UNPLANNED Area Radiation Monitor reading rise on ANY radiation monitors in Table R2.
Table R2 Fuel Handling Radiation Monitors R-2 Containment (Mode 6 and D)
R-5 Spent Fuel Pool (All Modes)
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-41 EP-AA-10XX (Rev X)
RU2 (cont)
Basis (cont):
This IC addresses a loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.
A water level loss will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available) or from any other temporarily installed monitoring instrumentation. A significant drop in the water level may also cause a rise in the radiation levels of adjacent areas that can be detected by monitors in those locations.
The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may rise due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with 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.
Alarm Response Procedure AR-K-29 SFP HI TEMP HI-LO LEVEL
- 3.
Technical Specifications Section 3.7.11 Spent Fuel Pool (SFP) Water Level
- 4.
Technical Specifications Section 3.9.6 Refueling Water Cavity Level
- 5.
P-9 Radiation Monitoring System
- 6.
CALC-2013-0001 Ginna Spent Fuel Area Dose Rates at Severely Reduced Water Levels
- 7.
66-9218244, Factory Acceptance Test Report of SFP Level Instrumentation System.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-42 EP-AA-10XX (Rev X)
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, 6, D Emergency Action Level (EAL):
Note:
If the equipment in the rooms or areas listed in Table R4 was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted.
- 1. Dose rate > 15 mR/hr in ANY of the areas contained in Table R3 Table R3 Areas Requiring Continuous Occupancy Main Control Room - R-1 Central Alarm Station - (by survey)
- 2. UNPLANNED event results in radiation levels that prevent or significantly impede access to ANY of the plant rooms or areas contained in Table R4.
Table R4 Areas with Entry-Related Mode Applicability Area Entry-Related Mode Applicability Auxiliary Building Top Floor Mode 3, 4, and 5 Auxiliary Building Middle Level Auxiliary Building Basement
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-43 EP-AA-10XX (Rev X)
RA3 (cont)
Basis:
UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal plant procedures. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR; therefore, Table R4 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.
This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room.
For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the elevated radiation levels preclude the ability to place shutdown cooling in service. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g.,
installing temporary shielding beyond that required by procedure, 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.).
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-44 EP-AA-10XX (Rev X)
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.
P-9 Radiation Monitoring System
- 3.
O-2 series of procedures
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS Month XXXX GN 2-45 EP-AA-10XX (Rev X)
RU3 Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
- 1. Letdown Monitor (R-9) reading 4.8 R/hr OR
EAL#1 Basis 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.
EAL #2 and 3 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.
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 Section 3.4.16 Reactor Coolant System-RCS Specific Activity
- 3.
CALC-2011-0019 NEI 99-01 Technical Basis for the Ginna R-9 Letdown Line Monitor Emergency Action Levels (EAL)
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-46 EP-AA-10XX (Rev X)
FG1 Initiating Condition:
Loss of ANY two barriers AND Loss or Potential Loss of the third barrier.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
Basis:
Fuel Cladding, RCS and Containment comprise the fission product barriers.
At the General Emergency classification level, each barrier is weighted equally.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-47 EP-AA-10XX (Rev X)
FS1 Initiating Condition:
Loss or Potential Loss of ANY two barriers.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
Basis:
Fuel Cladding, RCS and Containment comprise the fission product barriers.
At the Site Area Emergency classification level, each barrier is weighted equally.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-48 EP-AA-10XX (Rev X)
FA1 Initiating Condition:
ANY Loss or ANY Potential Loss of either Fuel Clad or RCS barrier.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.
Basis:
Fuel Cladding, RCS and Containment comprise the fission product barriers.
At the 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 the Containment barrier in combination with loss or potential loss of either the Fuel Cladding or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-49 EP-AA-10XX (Rev X)
FC1 Initiating Condition:
RCS or SG Tube Leakage Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
POTENTIAL LOSS Orange Path conditions exist, F-0.2 Core Cooling Basis:
There is no Loss threshold associated with RCS or SG Tube Leakage.
Potential Loss Threshold #1 Basis:
Indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
CSFST for F-0.2 Core Cooling
- 3.
FR-C.2 Response to Degraded Core Cooling
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-50 EP-AA-10XX (Rev X)
FC2 Initiating Condition:
Inadequate Heat Removal Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
LOSS
- 1. Red Path conditions exist, F-0.2 Core Cooling POTENTIAL LOSS
- 2. Orange Path conditions exist, F-0.2 Core Cooling OR
- 3. Red Path conditions exist, F-0.3 Heat Sink Basis:
Loss Threshold #1 Basis Indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier.
Potential Loss Threshold #2 Basis Indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur.
Potential Loss Threshold #3 Basis This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators by reducing total feed flow; during these conditions, classification using this threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS Barrier RC2 Potential Loss threshold; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and raise RCS pressure to the point where mass will be lost from the system.
Heat Sink - RED when heat sink is required, indicates the ultimate heat sink function is under extreme challenge.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-51 EP-AA-10XX (Rev X)
FC2 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
CSFST for F-0.2 Core Cooling
- 3.
FR-C.1 Response to Inadequate Core Cooling
- 4.
FR-C.2 Response to Degraded Core Cooling
- 5.
CSFST for F-0.3 Heat Sink
- 6.
FR-H.1 Response to Loss of Secondary Heat Sink
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-52 EP-AA-10XX (Rev X)
FC3 Initiating Condition:
Containment Radiation / RCS Activity Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
LOSS
- 1. Containment radiation monitor R-29/R-30 reading > 700 R/hr.
- 2. Letdown Monitor reading (R-9) > 24 R/hr with letdown in service.
- 3. Coolant activity as sampled > 300Ci/gm Dose Equivalent I-131.
Basis:
Loss Threshold #1 Basis The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier RC3 Loss 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.
Loss Threshold #2 and #3 Basis This threshold indicates that RCS radioactivity concentration is greater than 300 Ci/gm dose equivalent I-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 / Containment Radiation.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-53 EP-AA-10XX (Rev X)
FC3 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
P-9 Radiation Monitoring System
- 3.
EP-EAL-0712, Criteria for Choosing Containment Radiation Values Indicating: Loss of Fuel Clad and Potential Loss of Containment for Ginna Station
- 4.
CALC-2011-0019 NEI 99-01 Technical Basis for the Ginna R-9 Letdown Line Monitor Emergency Action Levels (EAL)
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-54 EP-AA-10XX (Rev X)
FC5 Initiating Condition:
Emergency Director Judgment.
Operating Mode Applicability:
1, 2, 3, 4 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 may 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-3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-55 EP-AA-10XX (Rev X)
RC1 Initiating Condition:
RCS or SG Tube Leakage Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
LOSS
- 1. Automatic or manual SI actuation is required by EITHER of the following:
- b. Steam Generator tube RUPTURE.
POTENTIAL LOSS
- 2. RCS leak rate > 50 gpm with letdown isolated.
- 3. Red Path conditions exist, F-0.4 Integrity Basis:
UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
Loss Threshold #1 Basis This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of Safety Injection (SI). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage), or outside of containment.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-56 EP-AA-10XX (Rev X)
RC1 (cont)
Basis (cont):
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier CT1 Loss threshold will also be met.
Potential Loss Threshold #2 Basis This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an SI actuation has not occurred. The threshold is met when RCS leak rate, excluding normal reductions in RCS inventory due to letdown and RCP seal leakoff, is > 50 gpm.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage), or outside of containment.
If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier CT1 Loss Threshold will also be met.
Potential Loss Threshold #3 Basis This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
RCS Integrity - RED indicates an extreme challenge to the safety function derived from appropriate instrument readings.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
E-0 Reactor Trip or Safety Injection
- 3.
E-3 Steam Generator Tube Rupture
- 4.
AP-RCS.1 Reactor Coolant Leak
- 5.
UFSAR Table 9.3.6 CVCS Performance Parameters
- 6.
UFSAR Section 9.3.4.2.2.2 Charging Pump Control
- 7.
CSFST for F-0.4 Integrity
- 8.
FR-P.1 Response to Imminent Pressurized Thermal Shock Condition
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-57 EP-AA-10XX (Rev X)
RC2 Initiating Condition:
Inadequate Heat Removal Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
POTENTIAL LOSS Red Path conditions exist, F-0.3 Heat Sink Basis:
There is no Loss threshold associated with Inadequate Heat Removal.
Potential Loss Threshold Basis Heat Sink - RED when heat sink is required indicates the ultimate heat sink function is under extreme challenge.
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators by reducing total feed flow; during these conditions, classification using this threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier FC2 Potential Loss threshold # 3; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and raise RCS pressure to the point where mass will be lost from the system.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
CSFST for F-0.3 Heat Sink
- 3.
FR-H.1 Response to Loss of Secondary Heat Sink
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-58 EP-AA-10XX (Rev X)
RC3 Initiating Condition:
Containment Radiation / RCS Activity Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
LOSS Containment radiation monitor R-29/R-30 reading > 10 R/hr.
Basis:
Loss Threshold Basis The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier FC3 Loss Threshold #1 since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
P-9 Radiation Monitoring System
- 3.
EP-EAL-0512, Criteria for Choosing Containment Radiation Monitor Readings Indicative of Loss of RCS for Ginna Station.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-59 EP-AA-10XX (Rev X)
RC5 Initiating Condition:
Emergency Director Judgment.
Operating Mode Applicability:
1, 2, 3, 4 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 may 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-3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-60 EP-AA-10XX (Rev X)
CT1 Initiating Condition:
RCS or SG Tube Leakage Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
LOSS A leaking or RUPTURED SG > 50gpm is FAULTED outside of containment.
Basis:
RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
Loss Threshold Basis This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier RC1 Potential Loss Threshold 2 and Loss Threshold 1.b, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC RU3 for the fuel clad barrier (i.e., RCS activity values) and IC MU6 for the RCS barrier (i.e., RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-61 EP-AA-10XX (Rev X)
CT1 (cont)
Basis (cont):
sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, gland seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
Primary-to-Secondary Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per MU6 Unusual Event per MU6 RCS leak rate > 50 gpm with letdown isolated (RCS Barrier Potential Loss)
Site Area Emergency per FS1 Alert per FA1 Requires an automatic or manual SI actuation (RCS Barrier Loss)
Site Area Emergency per FS1 Alert per FA1 There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-62 EP-AA-10XX (Rev X)
CT1 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
E-2 Faulted Steam Generator Isolation
- 3.
E-3 Steam Generator Tube Leakage
- 4.
ECA-1.2 LOCA Outside Containment
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-63 EP-AA-10XX (Rev X)
CT2 Initiating Condition:
Inadequate Heat Removal Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
POTENTIAL LOSS Red Path conditions exist, F-0.2 Core Cooling AND Functional Restoration procedures not effective in < 15 minutes.
Basis:
IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
There is no Loss threshold associated with Inadequate Heat Removal.
Potential Loss Threshold Basis This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered effective if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events.
Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
CSFST for F-0.2 Core Cooling
- 3.
FR-C.1 Response to Inadequate Core Cooling
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-64 EP-AA-10XX (Rev X)
CT3 Initiating Condition:
Containment Radiation / RCS Activity Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
POTENTIAL LOSS Containment radiation monitor R-29/R-30 reading > 7000 R/hr.
Basis:
There is no Loss threshold associated with Containment Radiation / RCS Activity.
Potential Loss Threshold Basis The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into 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 (FC3) and RCS Barrier Loss (RC3) 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-3
- 2.
P-9 Radiation Monitoring System
- 3.
EP-EAL-0712, Criteria for Choosing Containment Radiation Values Indicating: Loss of Fuel Clad and Potential Loss of Containment for Ginna Station
- 4.
NUREG-1228 Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-65 EP-AA-10XX (Rev X)
CT4 Initiating Condition:
Containment Integrity or Bypass Operating Mode Applicability:
1, 2, 3, 4 Fission Product Barrier (FPB) Threshold:
LOSS
- 1. Containment isolation is required AND EITHER of the following:
- a. UNPLANNED lowering in containment pressure or rise in radiation monitor readings outside of containment that in the Emergency Directors judgment indicate a loss of containment integrity.
- b. UNISOLABLE pathway from containment to the environment exists.
- 2. Indications of RCS leakage outside of containment POTENTIAL LOSS
- 3. Red Path conditions exist, F-0.5 Containment.
- 4. Hydrogen Concentration in Containment > 4%.
- 5. a. Containment pressure > 28 psig AND
- b. EITHER of the following conditions for > 15 minutes:
< 2 CRFC units operating
< 1 CS pump operating Basis:
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-66 EP-AA-10XX (Rev X)
CT4 (cont)
Basis (cont):
Loss Threshold #1 Basis These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both loss thresholds 1.a and 1.b.
1.a - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 3-F-1. Two simplified examples are provided.
One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
1.b - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term environment includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-67 EP-AA-10XX (Rev X)
CT4 (cont)
Basis (cont):
Refer to the top piping run of Figure 3-F-1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,
containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 3-F-1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met.
If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then Loss Threshold #2 would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 1.a to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using the Containment Barrier CT1 Loss Threshold.
Loss Threshold #2 Basis Containment sump, temperature, pressure and/or radiation levels will rise if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Rises in sump level, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not rise significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-68 EP-AA-10XX (Rev X)
CT4 (cont)
Basis (cont):
Refer to the middle piping run of Figure 3-F-1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause loss threshold 1.a to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Barrier RC1 Loss Threshold 1.a and/or Potential Loss threshold 2.a to be met.
Potential Loss Threshold #3 Basis Containment CSF RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings and/or sampling results, and thus represents a potential loss of containment.
If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Potential Loss Threshold #4 Basis The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.
Potential Loss Threshold #5 Basis This threshold describes a condition where containment pressure is greater than the set point at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The < 1 CS pump operating criteria is not considered met when CS pumps are purposely shut off in accordance with emergency operating procedure guidance (i.e., ES-1.3, Transfer to Cold Leg Recirculation). The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, containment recirculation fan coolers, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-69 EP-AA-10XX (Rev X)
CT4 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, Table 9-F-3
- 2.
UFSAR Figure 15.6-34 Containment Pressure Used for the R.E. Ginna Best-Estimate Large Break LOCA
- 3.
UFSAR Figures 6.2-1 through 6.2-6
- 4.
CSFST for F-0.5 Containment
- 5.
UFSAR 3.1.2.2.7 General Design Criterion 16 - Containment Design
- 6.
SAG-7, Reduce Containment Hydrogen
- 7.
UFSAR 1.5.10 Development of Hydrogen Recombiner
- 8.
UFSAR Section 6.2.2 Containment Heat Removal Systems
- 9.
UFSAR Section 6.2.1.2.3 Secondary System Pipe Break Analysis
- 10.
UFSAR Section 6.2.2.1.3 Design Evaluation
- 11.
Technical Specifications Section B 3.6 Containment Systems
- 12.
FR-Z.1 Response to High Containment Pressure
- 13.
ES-1.3, Transfer to Cold Leg Recirculation
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-70 EP-AA-10XX (Rev X)
Figure 3-F-1: PWR Containment Integrity or Bypass Examples Inside Containment Vent Auxiliary Building Filter Closed Cooling Water System RCP Seal Cooling Interface leakage point Open valve 1.a -
Airborne release from penetration Penetration Damper Damper Process Monitor Airborne Monitor Area Monitor Effluent Monitor 2 - RCS leakage outside CNMT Open valve Open valve Open valve Open valve Open valve 1.b - Airborne release from pathway 1.a -
Airborne release from valve Pump
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Month XXXX GN 2-71 EP-AA-10XX (Rev X)
CT5 Initiating Condition:
Emergency Director Judgment.
Operating Mode Applicability:
1, 2, 3, 4 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 may 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-3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-72 EP-AA-10XX (Rev X)
MG1 Initiating Condition:
Prolonged loss of all offsite and all onsite AC power to emergency buses.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. Loss of ALL offsite and ALL onsite AC power to 480V safeguards buses 14 and 16.
AND
- 2. EITHER of the following:
- a. Restoration of 480V safeguards bus 14 or 16 in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.
- b. Red Path conditions exist, F-0.2 Core Cooling.
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 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-73 EP-AA-10XX (Rev X)
MG1 (cont)
Basis (cont):
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.
Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SG1
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
- 3.
CSFST for F-0.2 Core Cooling
- 4.
FR-C.1 Response to Inadequate Core Cooling
- 5.
Ginna Station Blackout Program Section 3.7
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-74 EP-AA-10XX (Rev X)
MS1 Initiating Condition:
Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. Loss of ALL offsite and ALL onsite AC Power to 480V safeguards buses 14 and 16.
AND
- 2. Failure to restore power to 480V safeguards bus 14 or 16 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-75 EP-AA-10XX (Rev X)
MS1 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SS1
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
- 3.
ECA-0.0 Loss of All AC Power
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-76 EP-AA-10XX (Rev X)
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, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. AC power capability to 480V safeguards buses 14 and 16 reduced to only one of the following power sources for > 15 minutes.
Station Auxiliary Transformer 12A Station Auxiliary Transformer 12B Unit Auxiliary Transformer 11 backfeed Emergency Diesel Generator EDG 1A Emergency Diesel Generator EDG 1B AND
- 2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS powered from 480V safeguards buses 14 and 16.
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. Safeguards Buses 17 and 18 provide power to the four service water pumps and should not be considered in relation to EAL #2 because the availability of power to Buses 17 and 18 alone does not ensure engineered safety features required for hot shutdown and hot standby modes will be operable. 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 MU1.
An AC power source is a source recognized in AOPs and EOPs, and is 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).
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-77 EP-AA-10XX (Rev X)
MA1 (cont)
Basis (cont):
A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit auxiliary transformer 11.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level would be via IC MS1.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SA1
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-78 EP-AA-10XX (Rev X)
MU1 Initiating Condition:
Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Loss of ALL of the following offsite AC power capability to 480V safeguards buses 14 and 16 for > 15 minutes.
Station Auxiliary Transformer 12A Station Auxiliary Transformer 12B Unit Auxiliary Transformer 11 backfeed Basis:
This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses.
This condition represents a potential reduction in the level of safety of the plant.
For emergency classification purposes, capability means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level would be via IC MA1.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-79 EP-AA-10XX (Rev X)
MU1 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SU1
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-80 EP-AA-10XX (Rev X)
MG2 Initiating Condition:
Loss of all AC and Vital DC power sources for 15 minutes or longer.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. Loss of ALL offsite and ALL onsite AC power to 480V safeguards buses 14 and 16.
AND
- 2. Voltage is < 110.6 VDC on unit 125 VDC buses 1A and 1B.
AND
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 EAL conditions are met.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-81 EP-AA-10XX (Rev X)
MG2 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SG8
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
- 3.
ECA-0.0 Loss of All AC Power
- 4.
UFSAR Section 8.3.2 Direct Current Power Systems
- 5.
Technical Specifications Basis B.3.8.b
- 6.
DA-EE-99-047 125 VDC System Loads and Voltages
- 7.
Ginna Technical Evaluation ECP-17-000028
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-82 EP-AA-10XX (Rev X)
MS2 Initiating Condition:
Loss of all vital DC power for 15 minutes or longer.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Voltage is < 110.6 VDC on unit 125 VDC buses 1A and 1B 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.
UFSAR Section 8.3.2 Direct Current Power Systems
- 3.
Technical Specifications Basis B.3.8.b
- 4.
DA-EE-99-047 125 VDC System Loads and Voltages
- 5.
Ginna Technical Evaluation ECP-17-000028
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-83 EP-AA-10XX (Rev X)
MS3 Initiating Condition:
Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.
Operating Mode Applicability:
1, 2 Emergency Action Level (EAL):
- 1. Automatic or Manual Trip did not shutdown the reactor as indicated by Reactor Power > 5%.
AND
- 2. ALL manual and local actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 5%.
AND
- 3. EITHER of the following conditions exist:
- a. RED Path conditions exist, F-0.2 Core Cooling.
- b. RED Path conditions exist, F-0.3 Heat Sink.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator manual and local 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.
If Core Cooling Red Path or Heat Sink Red Path conditions exist prior to a successful reactor shutdown (i.e., < 5% reactor power) then entry is required.
In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators by reducing total feed flow; during these Heat Sink Red Path conditions classification using this threshold is not warranted.
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-84 EP-AA-10XX (Rev X)
MS3 (cont)
Basis (cont):
Escalation of the emergency classification level would be via IC MG2 or FG1.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SS5
- 2.
E-0 Reactor Trip or Safety Injection
- 3.
FR-S.1 Response to Reactor Restart/ATWS
- 4.
CSFST for F-0.1 Subcriticality
- 5.
CSFST for F-0.2 Core Cooling
- 6.
CSFST for F-0.3 Heat Sink
- 7.
P-1 Reactor Control and Protection System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-85 EP-AA-10XX (Rev X)
MA3 Initiating Condition:
Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability:
1, 2 Emergency Action Level (EAL):
Note:
A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
- 1. Automatic or Manual Trip did not shutdown the reactor as indicated by Reactor Power > 5%.
AND
- 2. Manual actions taken at the reactor control 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 trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control console 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 console, since this event entails a significant failure of the RPS.
A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (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 console.
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-86 EP-AA-10XX (Rev X)
MA3 (cont)
Basis (cont):
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.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SA5
- 2.
E-0 Reactor Trip or Safety Injection
- 3.
FR-S.1 Response to Reactor Restart/ATWS
- 4.
CSFST for F-0.1 Subcriticality
- 5.
P-1 Reactor Control and Protection System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-87 EP-AA-10XX (Rev X)
MU3 Initiating Condition:
Automatic or manual trip 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 Trip did not shutdown the reactor as indicated by Reactor Power
> 5%.
AND
- b. Subsequent manual action taken at the reactor control console is successful in shutting down the reactor as indicated by Reactor Power < 5%.
- 2. a. Manual Trip did not shutdown the reactor as indicated by Reactor Power > 5%.
AND
- b. EITHER of the following:
- 1. Subsequent manual action taken at the reactor control console is successful in shutting down the reactor as indicated by Reactor Power
< 5%.
- 2. Subsequent Automatic Trip is 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 trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control console or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
EAL #1 Basis Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control console to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-88 EP-AA-10XX (Rev X)
MU3 (cont)
Basis (cont):
heat generation will quickly fall to a level within the capabilities of the plants 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 console to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.
A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control console.
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control console 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.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied:
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 trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal generated as a result of plant work does not cause a plant transient but should have generated an RPS trip signal and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-89 EP-AA-10XX (Rev X)
MU3 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SU5
- 2.
E-0 Reactor Trip or Safety Injection
- 3.
FR-S.1 Response to Reactor Restart/ATWS
- 4.
CSFST for F-0.1 Subcriticality
- 5.
P-1 Reactor Control and Protection System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-90 EP-AA-10XX (Rev X)
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, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. a. UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for >15 minutes.
AND
- b. ANY Table M2 transient in progress.
Table M1 - Control Room Parameters Reactor Power PZR Level RCS Pressure Core Exit Temperature Level in at least one Steam Generator Steam Generator Auxiliary Feed Water Flow Table M2 - Significant Transients Automatic Turbine Runback >25% thermal reactor power Electrical Load Rejection >25% full electrical load Reactor Trip Safety Injection Actuation
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-91 EP-AA-10XX (Rev X)
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, 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 ICs FS1 or IC RS1.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SA2
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-92 EP-AA-10XX (Rev X)
MU4 Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the 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.
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, computer point, digital and recorder source within the Control Room.
Table M1 - Control Room Parameters Reactor Power PZR Level RCS Pressure Core Exit Temperature Level in at least one Steam Generator Steam Generator Auxiliary Feed Water Flow
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-93 EP-AA-10XX (Rev X)
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
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-94 EP-AA-10XX (Rev X)
MA5 Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.
Operating Mode Applicability:
1, 2, 3, 4 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.
- 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-95 EP-AA-10XX (Rev X)
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" means required to be operable by Technical Specifications for the current operating mode). This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. Manual or automatic electrical isolation of safety equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.
EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode and is in operation, since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
EAL #2.b addresses damage to a SAFETY SYSTEM component that is required to be operable by Technical Specifications for the current operating mode and is not in operation or readily apparent through indications alone, as well as damage to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.
Escalation of the emergency classification level would be via IC FS1 or RS1.
If the EAL conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.
Basis Reference(s):
- 1.
NEI 99-01, Rev 6 SA9
- 2.
ER-SC.4 Earthquake Emergency Plan
- 3.
ER-SC.1 Adverse Weather Plan
- 4.
ER-SC.2 High Water (Flood) Plan
- 5.
Ginna Station Fire Protection Program Volume 1 Part III Section 7 Fire Area/Fire Zone Analysis
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-96 EP-AA-10XX (Rev X)
MU6 Initiating Condition:
RCS leakage for 15 minutes or longer.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. RCS unidentified or pressure boundary leakage > 10 gpm for > 15 minutes.
- 2. RCS identified leakage >25 gpm for > 15 minutes.
- 3. Leakage from the RCS to a location outside containment >25 gpm for > 15 minutes.
Basis:
UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.
This IC addresses RCS leakage which may be a precursor to a more significant event.
In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.
EAL #1 and EAL #2 Basis These EALs are focused on a loss of mass from the RCS due to unidentified leakage",
"pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications).
EAL #3 Basis This EAL addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system.
These EALs thus apply to leakage into the containment, a secondary-side system (e.g.,
steam generator tube leakage) or a location outside of containment.
The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-97 EP-AA-10XX (Rev X)
MU6 (cont)
Basis (cont):
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level would be via ICs of Recognition Category R or F.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SU4
- 2.
Technical Specifications Section 3.4.13, RCS Operational Leakage
- 3.
AP-RCS.1 Reactor Coolant Leak
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-98 EP-AA-10XX (Rev X)
MU7 Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
- 1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine operations.
- 2. Loss of ALL Table M3 Offsite communications capability affecting the ability to perform offsite notifications.
- 3. Loss of ALL Table M3 NRC communications capability affecting the ability to perform NRC notifications.
Table M3 - Communications Capability System Onsite Offsite NRC Radios/Walkie Talkies X
Plant Page System X
Direct Dial POTS Lines (Blue Phones) System X
X X
Commercial Phone System X
X X
FTS 2001 telephone system (ENS, HPN)
X X
Control Room Hard Wired Satellite Phone X
X Control Room Emergency Cell Phone X
X RECS X
Basis:
This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-99 EP-AA-10XX (Rev X)
MU7 (cont)
Basis (cont):
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.
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-CE-114-100-F-07, GNP NY State Radiological Emergency Data Form (Part 1) (CNG).
EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SU6
- 2.
A-56 Communication Systems at Ginna Station
- 3.
ER-COMM.1 Loss of Communications
- 4.
EP-CE-114-100-F-07, GNP NY State Radiological Emergency Data Form (Part
- 1) (CNG).
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-100 EP-AA-10XX (Rev X)
MU8 Initiating Condition:
Failure to isolate containment or loss of containment pressure control.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Level (EAL):
- 1. a. Failure of containment to isolate when required by an actuation signal.
AND
- b. ANY required penetration remains open > 15 minutes after the actuation signal.
- 2. a. Containment pressure > 28 psig.
AND
- b. Either of the following conditions for > 15 minutes:
< 2 CRFC units operating
< 1 CS pump operating Basis:
This IC addresses a failure of any containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
EAL #1 Basis The containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL #2 Basis Addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Month XXXX GN 2-101 EP-AA-10XX (Rev X)
MU8 (cont)
Basis (cont):
indicates that containment heat removal/depressurization systems (e.g., containment sprays or containment recirculation fan coolers) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, SU7
- 2.
F-0.5 Containment
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-102 EP-AA-10XX (Rev X)
CA1 Initiating Condition:
Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability:
5, 6, D Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. Loss of ALL offsite and ALL onsite AC power to 480V safeguards buses 14 and
- 16.
AND
- 2. Failure to restore power to 480V safeguards bus 14 or 16 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-103 EP-AA-10XX (Rev X)
CA1 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, CA2
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-104 EP-AA-10XX (Rev X)
CU1 Initiating Condition:
Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability:
5, 6, D Emergency Action Level (EAL):
Note:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
- 1. AC power capability to 480V safeguards buses 14 and 16 reduced to only one of the following power sources for > 15 minutes.
Station Auxiliary Transformer 12A Station Auxiliary Transformer 12B Unit Auxiliary Transformer 11 backfeed Emergency Diesel Generator EDG 1A Emergency Diesel Generator EDG 1B AND
- 2. Any additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS powered from 480V safeguards buses 14 and 16.
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. Safeguards Buses 17 and 18 provide power to the four service water pumps and should not be considered in relation to EAL #2 because the availability of power to Buses 17 and 18 alone does not ensure engineered safety features required for hot shutdown and hot standby modes will be operable. 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
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-105 EP-AA-10XX (Rev X)
CU1 (cont)
Basis (cont):
modes, this condition is considered to be a potential degradation of the level of safety of the plant.
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 all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit auxiliary transformer 11.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA1.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6 CU2
- 2.
UFSAR Section 8 and Figure 8.1-1 Electrical Distribution System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-106 EP-AA-10XX (Rev X)
CA2 Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.
Operating Mode Applicability:
5, 6 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.
- 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-107 EP-AA-10XX (Rev X)
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" means 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
- 2.
ER-SC.4 Earthquake Emergency Plan
- 3.
ER-SC.1 Adverse Weather Plan
- 4.
ER-SC.2 High Water (Flood) Plan
- 5.
Ginna Station Fire Protection Program Volume 1 Part III Section 7 Fire Area/Fire Zone Analysis
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-108 EP-AA-10XX (Rev X)
CU3 Initiating Condition:
Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability:
5, 6 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 < 110.6 VDC on required unit 125 VDC buses1A and 1B 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-109 EP-AA-10XX (Rev X)
CU3 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, CU4
- 2.
UFSAR Section 8.3.2 Direct Current Power Systems
- 3.
Technical Specification Basis B.3.8.6 Battery Cell Parameters
- 4.
DA-EE-99-047 125 VDC System Loads and Voltages
- 5.
Ginna Technical Evaluation ECP-17-000028
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-110 EP-AA-10XX (Rev X)
CU4 Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
5, 6, D Emergency Action Level (EAL):
- 1. Loss of ALL Table C1 Onsite communications capability affecting the ability to perform routine operations.
- 2. Loss of ALL Table C1 Offsite communications capability affecting the ability to perform offsite notifications.
- 3. Loss of ALL Table C1 NRC communications capability affecting the ability to perform NRC notifications.
Table C1 - Communications Capability System Onsite Offsite NRC Radios/Walkie Talkies X
Plant Page System X
Direct Dial POTS Lines (Blue Phones) System X
X X
Commercial Phone System X
X X
FTS 2001 telephone system (ENS, HPN)
X X
Control Room Hard Wired Satellite Phone X
X Control Room Emergency Cell Phone X
X RECS X
Basis:
This IC addresses a significant loss of onsite, offsite, or NRC communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.
This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-111 EP-AA-10XX (Rev X)
CU4 (cont)
Basis (cont):
EAL #1 Basis Addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are listed in procedure EP-CE-114-100-F-07, GNP NY State Radiological Emergency Data Form (Part1) (CNG).
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.
A-56 Communication Systems at Ginna Station
- 3.
ER-COMM.1 Loss of Communications
- 4.
EP-CE-114-100-F-07, GNP NY State Radiological Emergency Data Form (Part1)
(CNG)
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-112 EP-AA-10XX (Rev X)
CA5 Initiating Condition:
Inability to maintain the plant in cold shutdown.
Operating Mode Applicability:
5, 6 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 to > 200ºF for > Table C2 duration.
- 2. UNPLANNED RCS pressure rise > 10 psig as a result of a temperature rise.
(This EAL does not apply in water-solid plant conditions.)
Table C2 - RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact Not Applicable 60 minutes*
Not Intact OR Reduced Inventory 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-113 EP-AA-10XX (Rev X)
CA5 (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 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 nozzle dams).
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, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). 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 or is at reduced inventory, 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 Basis Provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level would be via IC CS6 or RS1.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-114 EP-AA-10XX (Rev X)
CA5 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, CA3
- 2.
Technical Specifications Table 1.1-1
- 3.
O-2.2 Plant Shutdown from Hot Shutdown to Cold Conditions
- 4.
O-2.3.1A Containment Closure Capability Within Two Hours During RCS Reduced Inventory Operation
- 5.
O-2.3.1 Draining and Operation at Reduced Inventory of the Reactor Coolant System
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-115 EP-AA-10XX (Rev X)
CU5 Initiating Condition:
UNPLANNED rise in RCS temperature Operating Mode Applicability:
5, 6 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 to > 200ºF.
- 2. Loss of the following for >15 minutes.
ALL RCS temperature indications AND ALL RCS 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 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, which 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 nozzle dams).
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-116 EP-AA-10XX (Rev X)
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 Table C2 specific time criteria.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, CU3
- 2.
Technical Specifications Table 1.1-1
- 3.
O-2.2 Plant Shutdown from Hot Shutdown to Cold Conditions
- 4.
O-2.3.1A Containment Closure Capability Within Two Hours During RCS Reduced Inventory Operation
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-117 EP-AA-10XX (Rev X)
CG6 Initiating Condition:
Loss of Reactor Vessel / RCS inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability:
5, 6 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. Reactor Vessel / RCS 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 Erratic Source Range Neutron Monitor indication.
OR Containment Radiation R-29 or R-30 > 3 R/hr.
AND
- c. Any Containment Challenge Indication (Table C4).
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-118 EP-AA-10XX (Rev X)
CG6 (cont)
Emergency Action Level (EAL) (cont):
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.
CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
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 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.
Table C3 - Indications of RCS Leakage UNPLANNED Containment Sump A level rise*
UNPLANNED Containment Sump B level rise*
UNPLANNED Auxiliary Bldg. Sump/Tank level rise*
UNPLANNED RCDT level rise*
UNPLANNED rise in RCS makeup Observation of leakage or inventory loss
- Rise in level is attributed to a loss of Reactor Vessel/RCS inventory.
Table C4 - Containment Challenge Indications Hydrogen Concentration in Containment > 4%
UNPLANNED rise in containment pressure CONTAINMENT CLOSURE not established*
- If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-119 EP-AA-10XX (Rev X)
CG6 (cont)
Basis (cont):
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.
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 in Table C4 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 reactor vessel/RCS 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 reactor vessel/RCS.
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-120 EP-AA-10XX (Rev X)
CG6 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, CG1
- 2.
P-9 Radiation Monitoring System
- 3.
P-6 Precautions, Limitations and Setpoints Nuclear Instrumentation System
- 4.
RF-601 Fuel Handling Accident Instructions
- 5.
O-2.3 Draining the Reactor Coolant System to < 84 but > 64
- 6.
O-2.3.1 Draining and Operation at Reduced Inventory of the Reactor Coolant System.
- 7.
UFSAR 5.1.3.6 Monitoring Reactor Coolant Leakage
- 8.
UFSAR 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary
- 9.
O-2.3.1.A Containment Closure Capability Within Two Hours During Reduced Inventory Operation
- 10.
SAG-7 Reduce Containment Hydrogen
- 11.
EP-EAL-0502, Estimation of Radiation Monitor Readings Indicating Core Uncovery During Refueling Ginna Station.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-121 EP-AA-10XX (Rev X)
CS6 Initiating Condition:
Loss of Reactor Vessel / RCS inventory affecting core decay heat removal capability.
Operating Mode Applicability:
5, 6 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. Reactor Vessel / RCS 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 Erratic Source Range Neutron Monitor indication.
OR Containment radiation R-29 or R-30 > 3 R/hr.
Table C3 - Indications of RCS Leakage UNPLANNED Containment Sump A level rise*
UNPLANNED Containment Sump B level rise*
UNPLANNED Auxiliary Bldg. Sump/Tank level rise*
UNPLANNED RCDT level rise*
UNPLANNED rise in RCS makeup Observation of leakage or inventory loss
- Rise in level is attributed to a loss of Reactor Vessel / RCS inventory.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-122 EP-AA-10XX (Rev X)
CS6 (cont)
Emergency Action Level (EAL) (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.
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.
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 reactor vessel/RCS 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 reactor vessel/RCS.
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-123 EP-AA-10XX (Rev X)
CS6 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, CS1
- 2.
P-9 Radiation Monitoring System
- 3.
P-6 Precautions, Limitations and Setpoints Nuclear Instrumentation System
- 4.
RF-601 Fuel Handling Accident Instructions
- 5.
O-2.3 Draining the Reactor Coolant System to < 84 but > 64
- 6.
O-2.3.1 Draining and Operation at Reduced Inventory of the Reactor Coolant System.
- 7.
UFSAR 5.1.3.6 Monitoring Reactor Coolant Leakage
- 8.
UFSAR 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary
- 9.
EP-EAL-0502, Estimation of Radiation Monitor Readings Indicating Core Uncovery During Refueling Ginna Station.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-124 EP-AA-10XX (Rev X)
CA6 Initiating Condition:
Loss of Reactor Vessel / RCS inventory.
Operating Mode Applicability:
5, 6 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 Reactor Vessel / RCS inventory as indicated by RCS water level < 6 in. on Loop B indicator (LI-432B) OR Loop A compensated indication (LI-432A corrected).
- 2. a. Reactor Vessel / RCS level cannot be monitored for > 15 minutes.
AND
- b. Loss of Reactor Vessel / RCS inventory per Table C3 indications.
Table C3 - Indications of RCS Leakage UNPLANNED Containment Sump A level rise*
UNPLANNED Containment Sump B level rise*
UNPLANNED Auxiliary Bldg. Sump/Tank level rise*
UNPLANNED RCDT level rise*
UNPLANNED rise in RCS makeup Observation of leakage or inventory loss
- Rise in level is attributed to a loss of Reactor Vessel / RCS inventory.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-125 EP-AA-10XX (Rev X)
CA6 (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 conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.
EAL #1 Basis A lowering of water level below indicated RCS water level of < 6 in. on Loop B indicator or Loop A compensated indication indicates that operator actions have not been successful in restoring and maintaining reactor vessel/RCS water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery.
Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA5.
EAL #2 Basis The inability to monitor reactor vessel/RCS 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 reactor vessel/RCS.
The 15-minute duration for the loss of level indication was chosen because it is half of the Threshold duration specified in IC CS6 If the reactor vessel/RCS inventory 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.
RF-601 Fuel Handling Accident Instructions
- 3.
O-2.3 Draining the Reactor Coolant System to < 84 but > 64
- 4.
O-2.3.1 Draining and Operation at Reduced Inventory of the Reactor Coolant System.
- 5.
UFSAR 5.1.3.6 Monitoring Reactor Coolant Leakage
- 6.
UFSAR 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-126 EP-AA-10XX (Rev X)
CU6 Initiating Condition:
UNPLANNED loss of Reactor Vessel / RCS inventory for 15 minutes or longer.
Operating Mode Applicability:
5, 6 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 Reactor Vessel / RCS level to > procedurally established lower limit for > 15 minutes.
- 2. a. Reactor Vessel / RCS level cannot be monitored.
AND
- b. Loss of Reactor Vessel / RCS inventory per Table C3 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.
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 reactor vessel/RCS 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.
Table C3 - Indications of RCS Leakage UNPLANNED Containment Sump A level rise*
UNPLANNED Containment Sump B level rise*
UNPLANNED Auxiliary Bldg. Sump/Tank level rise*
UNPLANNED RCDT level rise*
UNPLANNED rise in RCS makeup Observation of leakage or inventory loss
- Rise in level is attributed to a loss of Reactor Vessel / RCS inventory.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Month XXXX GN 2-127 EP-AA-10XX (Rev X)
CU6 (cont)
Basis (cont):
Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.
EAL #1 Basis Recognizes that the minimum required reactor vessel/RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This Threshold 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 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.
The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.
EAL #2 Basis Addresses a condition where all means to determine reactor vessel/RCS 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 reactor vessel/RCS.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA6 or CA5.
Basis Reference(s):
- 1.
NEI 99-01, Rev. 6 CU1
- 2.
UFSAR 5.1.3.6 Monitoring Reactor Coolant Leakage
- 3.
UFSAR 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-128 EP-AA-10XX (Rev X)
HS1 Initiating Condition:
HOSTILE ACTION within the PROTECTED AREA.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-129 EP-AA-10XX (Rev X)
HS1 (cont)
Basis (cont):
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 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 HA1. 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 RG1, RG2 or HG7.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, HS1
- 2.
Ginna Safeguards Contingency Plan
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-130 EP-AA-10XX (Rev X)
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, 6, D Emergency Action Level (EAL):
- 1. A validated notification from NRC of an aircraft attack threat < 30 minutes from the site.
- 2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-131 EP-AA-10XX (Rev X)
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 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 ER-SEC.3, Response to Airborne Threat.
EAL #2 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-132 EP-AA-10XX (Rev X)
HA1 (cont)
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, HA1
- 2.
Ginna Safeguards Contingency Plan
- 3.
ER-SEC.3, Response to Airborne Threat
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-133 EP-AA-10XX (Rev X)
HU1 Initiating Condition:
Confirmed SECURITY CONDITION or threat.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, 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.
- 2. A validated notification from the NRC providing information of an aircraft threat.
- 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-134 EP-AA-10XX (Rev X)
HU1 (cont)
Basis (cont):
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.
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, Security Assessment and Response to Unusual Activities.
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 ER-SEC.3, Response to Airborne Threat.
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 HA1.
Basis Reference(s):
- 1.
NEI 99-01 Rev 6, HU1
- 2.
Ginna Safeguards Contingency Plan
- 3.
ER-SEC.1 Response to Change in Security Threat Level
- 4.
ER-SEC.2 Response to Intrusion by Adversary
- 5.
ER-SEC.3 Response to Airborne Threat
- 6.
SY-AA-101-132, Security Assessment and Response to Unusual Activities
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-135 EP-AA-10XX (Rev X)
HS2 Initiating Condition:
Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, 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 AP-CR.1 or the ER-FIRE series.
AND
- 2. Control of ANY Table H1 key safety function is not reestablished in < 35 minutes.
Table H1 - Safety Functions Reactivity Control (ability to shut down the reactor and keep it shutdown)
Core Cooling (ability to cool the core)
RCS Heat Removal (ability to maintain heat sink)
Basis:
The time period to establish control of the plant starts when either:
- a.
Control of the plant is no longer maintained in the Main Control Room OR
- b.
The last Operator has left the Main Control Room.
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant 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 35 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-136 EP-AA-10XX (Rev X)
HS2 (cont)
Basis (cont):
Escalation of the emergency classification level would be via IC FG1 or CG6.
Basis Reference(s):
- 1.
NEI 99-01, Rev 6 HS6
- 2.
AP-CR.1 Control Room Inaccessibility
- 3.
OP-GI-102-106, Operator Response Time Program at Ginna
- 4.
ER-FIRE series of procedures
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-137 EP-AA-10XX (Rev X)
HA2 Initiating Condition:
Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per AP-CR.1 or the ER-FIRE series.
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.
AP-CR.1 Control Room Inaccessibility
- 3.
ER-FIRE series of procedures
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-138 EP-AA-10XX (Rev X)
HU3 Initiating Condition:
FIRE potentially degrading the level of safety of the plant.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, 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.
Potential 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 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.
- 3. A FIRE within the plant PROTECTED AREA not extinguished in < 60 minutes of the initial report, alarm or indication.
- 4. A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
Table H2 - Vital Areas Reactor Containment Building Auxiliary Building Control Building Intermediate Building Emergency Diesel Buildings SAFW Building Screenhouse Cable Tunnel Battery Rooms
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-139 EP-AA-10XX (Rev X)
HU3 (cont)
Basis:
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.
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 alarms, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarms, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarms, indication or report.
EAL #2 Basis This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.
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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-140 EP-AA-10XX (Rev X)
HU3 (cont)
Basis (cont):
EAL #4 Basis If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.
Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
ISFSI is not specifically addressed in EAL #3 and #4 since it is within the plant PROTECTED AREA and is therefore covered under EALs #3 and #4.
Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA2 or MA5.
Basis Reference(s):
- 1.
NEI 99-01, Rev 6 HU4
- 2.
UFSAR Table 3.2-1 Classification of Structures, Systems and Components
- 3.
Ginna Station Fire Protection Program Volume I Part III Section 7.0 Fire Area/Fire Zone
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-141 EP-AA-10XX (Rev X)
HU4 Initiating Condition:
Seismic event greater than OBE levels.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
Note:
Potential escalation of the emergency classification level would be via IC CA2 or MA5.
For emergency classification if EAL #2 first three bullets are 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.
- 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.
If the above bullets are 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.
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 public will remain functional.
2 An SSE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-142 EP-AA-10XX (Rev X)
HU4 (cont)
Basis (cont):
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 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.
UFSAR Section 3.7.1.2 Design Response Spectra
- 3.
ER-SC.4 Earthquake Emergency Plan
- 4.
US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-143 EP-AA-10XX (Rev X)
HA5 Initiating Condition:
Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
3, 4, 5 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 ANY Table H3 area.
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.
Table H3 Areas with Entry Related Mode Applicability Area Entry Related Mode Applicability Auxiliary Building Top Floor Mode 3, 4 and 5 Auxiliary Building Middle Level Auxiliary Building Basement
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-144 EP-AA-10XX (Rev X)
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 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 precludes 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 Directors 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.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-145 EP-AA-10XX (Rev X)
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 or that automatically or manually activate a fire suppression system in an area.
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.
O-2 series of procedures
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-146 EP-AA-10XX (Rev X)
HU6 Initiating Condition:
Hazardous Event Operating Mode Applicability:
1, 2, 3, 4, 5, 6, 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.
Potential escalation of the emergency classification level would be via IC CA2 or MA5.
- 1. Tornado strike within the PROTECTED AREA.
- 2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode.
- 3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
- 4. A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
- 5. Lake level > 252 ft.
- 6. Screenhouse Suction Bay water level <19 ft or <17.5 ft by manual measurement.
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.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-147 EP-AA-10XX (Rev X)
HU6 (cont)
Basis (cont):
EAL #1 Basis Addresses a tornado striking (touching down) within the PROTECTED AREA.
EAL #2 Basis Addresses internal room or area flooding of a building 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 to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 Basis Ginna plant grade is generally at 270 ft mean sea level (msl) except the area between the lake and Turbine Building which is at grade 253 ft msl. Lake water level > 253.28 ft msl corresponds to plant design levels. A lake level of 252 ft has been selected for this threshold to be anticipatory of exceeding design flood levels and is the level at which flood control actions are procedurally taken.
High lake level may be determined using markers attached to a metal pole mounted on the discharge canal bridge upstream of the submarine net. The high level markers are at lake levels of 252 ft and 253 ft
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-148 EP-AA-10XX (Rev X)
HU6 (cont)
Basis (cont):
EAL #6 Basis The Screenhouse Lo-Lo level alarm actuates at 19' indicated and increased Control Room monitoring is initiated. This level has been selected for this threshold to be anticipatory of a potential loss of service water system pump suction at 16.0 ft. (14.5 feet using manual measurement)
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 3.3.2.1.4 Wind and Tornado Loadings - Input Load Criteria
- 3.
UFSAR Table 3.2-1 Classification of Structures, Systems and Components
- 4.
Ginna Station Fire Protection Program Volume I Part III Section 7.0 Fire Area/Fire Zone Analysis
- 5.
UFSAR Section 3.4.1 Flood Protection
- 6.
ER-SC.2 High Water (Flood) Plan
- 7.
AR-I-9 Screen House Lo-Lo Level 19
- 8.
ER-SC.3 Low Screenhouse Water Level
- 9.
ER-SC.1 Adverse Weather Plan
- 10.
UFSAR Section 13.5.2.2.3
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-149 EP-AA-10XX (Rev X)
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, 6, 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
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-150 EP-AA-10XX (Rev X)
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, 6, 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
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-151 EP-AA-10XX (Rev X)
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, 6, 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
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Month XXXX GN 2-152 EP-AA-10XX (Rev X)
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, 6, 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
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS Month XXXX GN 2-153 EP-AA-10XX (Rev X)
E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability:
1, 2, 3, 4, 5, 6, D Emergency Action Level (EAL):
Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:
> 1600 mRem/hr on the Horizontal Storage Module (HSM) front surface.
> 400 mRem/hr on the Horizontal Storage Module (HSM) door centerline OR
> 16 mRem/hr on the end shield wall exterior 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 x, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSIs are covered under ICs HU1, HA1, and HS1.
Ginna Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS Month XXXX GN 2-154 EP-AA-10XX (Rev X)
E-HU1 (cont)
Basis Reference(s):
- 1.
NEI 99-01, Rev 6 E-HU1
- 2.
Attachment A, Technical Specifications, TRANSNUCLEAR, INC.
Standardized NUHOMS Horizontal Modular Storage System, Certificate of Compliance NO. 1004, Amendment NO. 10, Docket 72-1004