ML19246A718

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Attachment 1 - 10 CFR 50.54(q)(5) Change Summary Analysis
ML19246A718
Person / Time
Site: FitzPatrick  Constellation icon.png
Issue date: 08/23/2019
From:
Exelon Generation Co
To:
Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
JAFP-19-0078, SFGL-12
Download: ML19246A718 (352)
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Text

ATTACHMENT 1 10 CFR 50.54(q)(5) Change Summary Analysis Change Summary Analysis Page 1 of 5 ATTACHMENT 1 10 CFR 50.54(q)(5) Change Summary Analysis Document Titles Exelon Generation Company, LLC (Exelon) has issued the following Emergency Plan document revisions for the James A. FitzPatrick Nuclear Power Plant (JAF):

  • EP-AA-1014, Addendum 3, Revision 1, "James A. FitzPatrick Nuclear Power Plant Emergency Action Levels"
  • EP-AA-1014, Addendum 3, Appendix 1, Revision 0, "JAF EAL Wallboard"
  • EP-AA-1014 Addendum 3, Revision 2, "James A. FitzPatrick Nuclear Power Plant Emergency Action Levels"
  • EP-AA-1014, Addendum 3, Appendix 1, Revision 1, "JAF EAL Wallboard" As a result of implementing EP-AA-1014, Addendum 3, Appendix 1, Revision 0, JAF emergency planning Procedure IAP-2, Figure IAP-2.1, "EAL Classification Matrix," has been superseded.

Description of Procedures EP-AA-1014, Addendum 3, along with the supporting Appendix 1 wallboard, describe the Emergency Action Levels (EALs) implemented at JAF for entering Emergency Classification Levels (ECLs).

Description of Changes EP-AA-1014, Addendum 3, Revision 1 (including Appendix 1. Revision 0)

JAF has implemented the revised Emergency Action Level (EAL) scheme based on the Nuclear Energy lnstitute's (NEl's) guidance document NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors." The NEI 99-01, Revision 6, EAL scheme has been endorsed by the U.S. Nuclear Regulatory Commission (NRC).

By letter dated January 31, 2018 (ML18037A782), as supplemented by letter dated July 12, 2018 (ML18194A252), Exelon submitted a License Amendment Request (LAR) to the NRC requesting adoption of the NEI 99-01, Revision 6, EAL scheme for JAF. Subsequently, as documented in an NRC letter and supporting Safety Evaluation Report (SER) dated November 28, 2018 (ML18289A432), the NRC issued Amendment No. 323 to Renewed Facility Operating License (RFOL) No. DPR-59 for JAF approving the LAR supporting the implementation of the NEI 99-01, Revision 6, EAL scheme for JAF.

The NRG-approved NEI 99-01, Revision 6, EAL scheme for JAF is being implemented as EP-AA-1014 Addendum 3, Revision 1. In conjunction with the implementation of this revision to EP-AA-1014, Addendum 3, JAF has also implemented Appendix 1 to Addendum 3 to include the revised EAL Wallboard. The revised wallboard EP-AA-1014 Addendum 3, Appendix 1 Revision 0, replaces Procedure IAP-2, Figure IAP-2.1, "EAL Classification Matrix."

Change Summary Analysis Page 2 of 5 EP-AA-1014, Addendum 3, Revision 2 (including Appendix 1, Revision 1)

The following changes were made to EP-AA-1014, Addendum 3, and the supporting Appendix, pertaining to EAL E-HU1 threshold values and EAL HU3 Basis section:

  • EAL E-HU1 NEI 99-01, Revision 6, provides the following information for the development of EAL E-HU1, "Independent Spent Fuel Storage Installation (ISFSI) Damage, "to a loaded cask CONFINEMENT BOUNDARY:

ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: All Example Emergency Action Levels:

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

Basis:

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 issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times'~ which is also used in Recognition Category A IC AU1, 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 /SFS/s are covered under /Cs HU1 and HA 1.

Developer Notes:

The results of the ISFSI Safety Analysis Report (SAR) [per NUREG 1536}, or a SAR referenced in the cask Certificate of Compliance and the related NRG

. Safety Evaluation Report, identify the natural phenomena events and accident Change Summary Analysis Page 3 of 5 conditions that could potentially affect the CONFINEMENT BOUNDARY. This EAL addresses damage that could result from the range of identified natural or man-made events (e.g., a dropped or tipped over cask, EXPLOSION, FIRE, EARTHQUAKE, etc.).

The allowable radiation level for a spent fuel cask can be found in the cask's technical specification located in the Certificate of Compliance.

EGL Assignment Attributes: 3. 1. 1.B JAF has installed the HI-STORM 100 Cask System. The governing Technical Specifications (TS) are in Appendix A of Certificate of Compliance (CoC) No. 1014, Amendment Nos. 0, 1, 2, 5, and 8. The current EAL thresholds encompass the casks in use under these Amendments. Additional casks are being added (i.e., HI-STORM Cask Nos. 1172, 1173, 1174, 1175, 1176, and 1177) and will be in use under the Amendment No. 8 thresholds.

Section 5.7.4 of the Appendix A TS has the following allowable peak dose rates for a HI-STORM 100 Cask System loaded overpack governed by CoC No. 1014, AmendfTlent No. 8, Appendix A:

o 30 mrem/hr (gamma + neutron) on the top of the overpack o 300 mrem/hr (gamma + neutron) on the side of the overpack excluding inlet and outlet ducts The NEI 99-01, Revision 6, guidance specifies using a value of two (2)"times the allowable radiation level as the EAL threshold basis. JAF EAL E-HU1 has been revised as follows to include thresholds for the use of the HI-STORM 100 Cask System governed by CoC No.

1014, Amendment No. 8, Appendix A:

o 60 mrem/hr (gamma + neutron) on the top of the spent fuel cask o 600 mrem/hr (gamma + neutron) on the side of the spent fuel cask excluding inlet and outlet ducts Therefore, EP-AA-1014, Addendum 3, and supporting Appendix 1 (i.e., EAL Wallboard) were revised to reflect this change.

  • EAL HU3 Additional clarification was added to the Basis section for EAL HU3. Specifically, for Table H2 in the Basis section, clarifying information was added to the: 1) bullet for Battery Rooms /

Battery Room Corridor, and 2) discussion for Battery Charger Rooms. This information was added to further clarify that a fire in a Battery Charger Room should be evaluated based on its effects on the 125 VDC Bus distribution against EALs MS2 (i.e., loss of all vital DC power for 15 minutes or longer when in hot plant operating modes), and EAL CU3 (i.e., loss of vital DC power for 15 minutes or longer when in cold plant operating modes). This clarification will help to avoid any Operator confusion and ensure the Operators have information readily available in support of making timely and accurate assessments of the threshold values for EAL entry conditions.

Change Summary Analysis Page 4 of 5

  • Description of How the Changes Still Comply with Regulations EP-AA-1014, Addendum 3, Revision 1 (including Appendix 1, Revision 0)

The change to the EAL scheme reflects the NRC's approval supporting the implementation of the NEI 99-01, Revision 6, EAL scheme for JAF. As such, the NRC determined that the changes meet existing applicable regulatory requirements and commitments.

In addition, replacing the current/superseding Figure IAP-2.1 in Procedure IAP-2 with the EAL wallboard in EP-AA-1014, Addendum 3, Appendix 1, which reflects the revised NEI 99-01, Revision 6, EAL scheme as approved, does not alter the meaning or intent of the approved EALs.

EP-AA-1014, Addendum 3, Revision 2 (including Appendix 1, Revision 1)

  • EAL E-HU1 The change to EAL E-HU1 reflects the use of additional HI-STORM 100 Cask System Overpacks at the Independent Spent Fuel Storage Installation (ISFSI) at JAF; specifically, HI-STORM Overpack Serial Nos. 1172, 1173, 1174, 1175, 1176, and 1177. The Coe No.

1014, Amendment No. 8, Appendix A TS documents the allowable peak dose rates for the additional HI-STORM 100 Cask System Overpacks. NEI 99-01, Revision 6, provides the guidance for the development of the EAL E-HU1 threshold values as two (2) times the allowable peak dose rates. As a result, this adds additional thresholds values of > 60 mrem/hr (gamma+ neutron) on top of the spent fuel cask, and> 600 mrem/hr (gamma+

neutron) on the side of the spent fuel cask, excluding inlet and outlet ducts. Updating the EAL threshold values based on an approved Coe does not alter the meaning or intent of the basis of the approved EAL.

.* EAL HU3 The change to Table H2 in the Basis Section for EAL HU3 to add further clarification as discussed above, does not alter the meaning or intent of the basis of the approved EAL.

The changes ensure that information is readily available to Operators to make timely and accurate assessments of the threshold values for EAL entry conditions. The changes are in keeping with the guidance specified in NEI 99-01, Revision 6. No existing emergency planning requirements have been deleted or minimized and applicable regulations and commitments to the NRC continue to be met.

Description of Why the Changes are Not a Reduction in Effectiveness (RIE)

The revisions to the Emergency Plan documents discussed above do not change the meaning or intent of Emergency Plan requirements for JAF. Existing emergency planning requirements and capabilities have not been deleted or reduced as a result of these revisions. The changes do not alter the capability of the Emergency Response Organization (ERO) to implement required Emergency Plan actions, and do not affect the timeliness or effectiveness of the Change Summary Analysis Page 5 of 5 performance of emergency planning functions. The changes described continue to satisfy applicable emergency planning standard requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E, as well as program elements described in the guidance of NUREG-0654.

The changes described continue to maintain emergency planning commitments to the NRC.

Therefore, the revisions to the Emergency Plan documents described above do not constitute a reduction in effectiveness of the Emergency Plan for JAF.

ATTACHMENT 2 Radiological Emergency Plan Document Revision EP-AA-1014, Addendum 3, Revision 1, "James A. FitzPatrick Nuclear Power Plant Emergency Action Levels"

EP-AA-1014 Addendum 3

~* Exelon Generations Revision 1 JAMES A. FITZPATRICK

  • NUCLEAR POWER PLANT EMERGENCY ACTION LEVELS I.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • REVISION HISTORY Rev. 1 Julv 2019

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • Section 1: Classification of Emergencies 1.1 General Section D of the Exelon Nuclear Standardized Emergency Plan divides the types of emergencies into four Emergency Classification Levels (ECLs). The four are the Unusual Event (UE), Alert, Site Area Emergency (SAE), and General Emergency (GE). These ECLs are entered by satisfying the Initiating Condition (IC) through meeting an Emergency Action Level (EAL) of the IC provided in this section of the Annex. The ECLs are escalated from least severe to most severe according to relative threat to the health and safety of the public and emergency workers. Depending on the severity of an event, prior to returning to a standard day-to-day organization, a state or phase called Recovery may be entered to provide dedicated resources and organization in support of restoration and communication activities following the termination of the emergency.

Unusual Event (UE): 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.

Alert: 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.

Site Area Emergency (SAE): Events are in progress or have occurred which involve an 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.

General Emergency (GE): 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 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • Recovery: Recovery can be* considered as a phase of the emergency and is entered by meeting emergency termination criteria provided in EP-CE-111 Emergency Classification and Protective Action Recommendations.

Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The Emergency Classification Levels, in ascending order of severity, are:

  • Unusual Event (UE)
  • Alert
  • Site Area Emergency (SAE)
  • General Emergency (GE)

Initiating Condition (IC): An event or condition that aligns with the definition of one of the four Emergency Classification Levels by virtue of the potential or actual effects or consequences.

Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given Emergency Classification Level.

An emergency is classified by assessing plant conditions and comparing abnormal conditions to ICs and EALs, based on the designated Operational Condition (MODE). Modes 1 through 5 are based on Reactor Mode Switch Position and average reactor coolant tempe.rature.

"Defueled" Mode was established for classification purposes under NEI 99-01 to reflect conditions where all fuel has been removed from the Reactor Pressure Vessel.

OPERATING MODES REACTOR MODE SWITCH POSITION TEMP (1) Power Operation: Run N/A (2) Startup: Refuel (a) or Startup/Hot Standby N/A (3) Hot Shutdown (a): Shutdown > 212° F (4) Cold Shutdown (a): Shutdown  ::; 212° F (5) Refueling (bl: Shutdown or Refuel N/A (D) Defueled: All reactor fuel removed from reactor pressure vessel (full core off load during refueling or extended outage).

(a> All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

Hot Matrix - applies in modes (1 ), (2), and (3)

Cold Matrix - applies in modes (4), (5), and (D)

July 2019 JAF 1-2 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear Individuals responsible for the classification of events will refer to the IC's and EALs on the

  • matrix of the appropriate station Standardized Emergency Plan Annex (this document). This matrix will contain ICs, EALs, Mode Applicability Designators, appropriate EAL numbering system, and additional guidance necessary to classify events. It may be provided as a user aid.

The matrix is set up in six Recognition Categories. The first is designated as "R" and relates to Abnormal Radiological Conditions / Abnormal Radiological Effluent Releases. The second is designated as "F" and relates to Fission Product Barrier Degradation. The third is designated as "M" and relates to hot condition System Malfunctions. The fourth is designated as "C" and relates to Cold Shutdown I Refueling System Malfunctions. The fifth is designated as "H" and relates to Hazards and Other Conditions Affecting Plant Safety. The sixth is designated "E-H" and relates to ISFSI Malfunctions.

The matrix is designed to provide an evaluation of the Initiating Conditions from the worst conditions (General Emergencies) on the left to the relatively less severe conditions on the right (Unusual Events). Evaluating conditions from left to right will reduce the possibility that an event will be under classified. All Recognition Categories should be reviewed for applicability prior to classification.

The Initiating Conditions are coded with a two letter and one number code. The first letter is the Recognition Category designator, the second letter is the classification Level, "U" for (Notification Of) Unusual Event, "A" for Alert, "S" for Site Area Emergency and "G" for General Emergency. The EAL number is a sequential number for that Recognition Category series. All ICs that are describing the severity of a common condition (series) will have the same number .

The EAL number may then be used to reference a corresponding page(s), which provides the basis information pertaining to the IC:

  • Mode Applicability
  • Basis Classification is not to be made without referencing, comparing and satisfying the specified Emergency Action Levels.

A list of definitions is provided as part of this document (Section 1.7) for terms having specific meaning to the EALs. Site specific definitions are provided for terms with the intent to be used for a particular IC/EAL and may not be applicable to other uses of that term at other sites, the Emergency Plan or procedures.

References are also included to documents that were used to develop the EALs .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear References to the Emergency Director means the person in Command and Control as defined

  • in the Emergency Plan. Classification of emergencies is a non-delegable responsibility of Command and Control for the onsite facilities with responsibility assigned to the Shift Emergency Director (Control Room Shift Manager) or the Site Emergency Director (Technical Support Center). Classification of emergencies remains the responsibility of the Shift Emergency Director until Command and Control is transferred to the Site Emergency Director (Technical Support Center).

Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL is IMMINENT. If, in the judgment of the Emergency Director, an IMMINENT situation is at hand, the classification should be made as if the EAL has been exceeded. While this is particularly prudent at the higher EGL (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all ECLs.

1.2 Classification, Instrumentation and Transient Events All classifications are to be based upon valid indications, reports or conditions. Indications, reports or conditions are considered valid when they are verified by (1) an instrument channel check, or (2) indications on related or redundant indications, or (3) by direct observation by plant personnel, such that doubt related to the indication's operability, the condition's existence, or the report's accuracy is removed. Implicit in this is the need for timely assessment.

  • Indications used for monitoring and evaluation of plant conditions include the normally used instrumentation, backup or redundant instrumentation, and the use of other parameters that provide information that supports determination if an EAL has been reached. When an EAL refers to a specific instrument or indication that is determined to be inaccurate or unavailable, then alternate indications shall be used to monitor the specified condition.

During an event that results in changing parameters trending towards an EAL classification, and instrumentation that was available to monitor this parameter becomes unavailable or the parameter goes off scale, the parameter should be assumed to have been exceeded consistent with the trend and the classification made if there are no other direct or indirect means available to determine if the EAL has not been exceeded.

The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the EAL to be exceeded (i.e., this is the time that the EAL information is first available) .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear Planned evolutions involve preplanning to address the limitations imposed by the condition, the

  • performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition in accordance with the specific requirements of the site's Technical Specifications. Activities which cause the site to operate beyond that allowed by the site's Technical Specifications, planned or unplanned, may result in an EAL being met or exceeded. Planned evolutions to test, manipulate, repair, perform maintenance or modifications to systems and equipment that result in an EAL being met or exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the specific operating license.

However, these conditions may be subject to the reporting requirements of 10 CFR 50. 72.

When two or more EALs are determined, declaration will be made on the highest classification level.

Concerning ECL Downgrading, Exelon Nuclear policy is that ECLs shall not be downgraded to a lower classification. Once declared, the event shall remain in effect until no classification is warranted or until such time as conditions warrant classification to Recovery.

There may be cases in which a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review), and the condition no longer exists. In these cases, an emergency should not be declared. Reporting requirements of 10 CFR 50.72 are applicable, the guidance of NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73 and the Reportability Reference Manual, should be applied.

.1.3 Mode Applicability The plant-operating mode that existed at the time that the event occurred, prior to any protective system or operator action initiated in response to the condition, is compared to the mode applicability of the EALs. If an event occurs, and a lower or higher plant-operating mode is reached before the emergency classification can be made, the declaration shall be based on the mode that existed at the time the event occurred.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that have Cold Shutdown or Refueling for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the Fission Product Barrier Matrix EALs are applicable only to events that initiate in Hot Shutdown or higher.

If there is a change in Mode following an event declaration, any subsequent events involving EALs outside of the current declaration escalation path will be evaluated on the Mode of the plant at the time the subsequent events occur.

1.4 Emergency Director Judgment Emergency Director (ED) Judgment EALs are provided in the Hazards and Other Condition Affecting Plant Safety section and on the Fission Product Barrier (FPB) Matrix. Both of the ED Judgment EALs have specific criteria for when they should be applied .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear The Hazards Section ED Judgment EALs are intended to address unanticipated conditions

  • which are not addressed explicitly by other EALs but warrant declaration of an emergency because conditions exist which are believed by the ED to fall under specific emergency classifications (UE, Alert, SAE or GE).

The FPS Matrix ED Judgment EALs are intended to include unanticipated conditions, which are not addressed explicitly by any of the other FPS threshold values, but warrant determination because conditions exist that fall under the broader definition for a significant Loss or Potential Loss of the fission product barrier (equal to or greater than the defined FPS threshold values).

1.5 Fission Product Barrier (FPB) Threshold A fission product barrier threshold is a pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

FPS thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary FPBs are:

Fuel Clad (FC)

Reactor Coolant System (RCS)

Containment (CT)

  • Upon determination that one or more FPS thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the FPS IC/EAL criteria to determine the appropriate ECL.

In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/

Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g.,

spent fuel pool accidents, design containment leakage following a LOCA, etc.).

1.6 Fission Product Barrier Restoration Fission Product Barriers are not treated the same as EAL threshold values. Conditions warranting declaration of the loss or potential loss of a FPS may occur resulting in a specific classification. The condition that caused the loss or potential loss declaration could be rectified as the result of Operator action, automatic actions, or designed plant response. Barriers will be considered re-established when there are direct verifiable indications (containment penetration or open valve has been isolated, coolant sample results, etc) that the barrier has been restored and is capable of mitigating future events .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear The reestablishment of a FPB does not alter or lower the existing classification. Termination

  • 1.7 and entry into Recovery phase is still required for exiting the present classification. However the reestablishment of the barrier should be considered in determining future classifications should plant conditions or events change.

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

CONTAINMENT CLOSURE: The procedurally defined actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

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.

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fire. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be

  • met by the station.

HOSTILE ACTION: An act toward a Nuclear Power Plant (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).

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.

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.

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.

NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value .

July 2019 JAF 1-7 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear 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 .

PROJECTILE: An object directed toward a Nuclear Power Plant (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.

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.

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.

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.

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 .

VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train 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 SAFETY SYSTEM train .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear EMERGENCY ACTION LEVEL TECHNICAL BASIS PAGE INDEX

  • EAL General RG1 Pg.

2-28 RG2 2-38 EAL Site Area RS1 RS2 2-39 Pg.

2-30 EAL RA1 RA2 RA3 Alert Pg.

2-32 2-40 2-42 Unusual EAL RU1 RU2 RU3 Event Pg.

2-35 2-43 2-45 FG1 2-49 FS1 2-50 FA1 2-51 Fuel Clad RCS Containment FC1 2-52 FC2 2-53 RC2 2-57 CT2 2-65 RC3 2-59 CT3 2-66 RC4 2-60 FC5 2-55 RC5 2-63 CT5 2-68 CT6 2-69 FC7 2-56 RC? 2-64 CT7 2-72 MG1 2-73 MS1 2-75 MA1 2-77 MU1 2-79 MG2 2-80 MS2 2-82 MS3 2-83 MA3 2-85 MU3 2-87 MA4 2-90 MU4 2-93 MA5 2-95

  • CA1 CA2 2-102 2-106 MU6 MU7 CU1 CU3 2-98 2-100 2-104 2-109 CU4 2-111 CA5 2-113 CU5 2-115 CG6 2-117 CS6 2-121 CA6 2-125 CU6 2-127 HS1 2-130 HA1 2-132 HU1 2-135 HS2 2-137 HA2 2-139 HU3 2-140 HU4 2-144 HA5 2-147 HU6 2-150 HG7 2-153 HS7 2-154 HA7 2-155 HU7 2-156 E-HU1 2-157

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Abnormal Rad Levels/ Radiological Effluents RG1 II]~@l@J~[g RS1 II]~@l@J~[g RA1 II]~@l@J~[g RU1 II]~@l@J~[g Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite Release of gaseous or liquid radioactivity greater than 2 times than 1000 mRem TEDE or 5000 mRem thyroid CDE. than 100 mRem TEDE or 500 mRem thyroid CDE. dose greater than 10 mRem TEDE or 50 mRem thyroid CDE. the ODCM limits for 60 minutes or longer.

Emergency Action Level (EAL): Emergency Action Level (EAL}: Emergency Action Level (EAL): Emergency Action Level (EAL):

Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has been Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has been Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has

. 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, or will likely be exceeded.

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

. 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

. exceeded 15 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the

. exceeded 15 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If

. exceeded 60 minutes.

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

.l!l C

Cl)

. purposes.

The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency

. purposes.

The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency

. classification purposes.

The pre-calculated effluent monitor values presented in EAL #1 (Table R 1) should be used for emergency classification purposes.

=

w iii classification assessments until the results from a dose assessment using actual meteorology are available.

classification assessments until the results from a dose assessment using actual m9teoro1ogy are available.

1.

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

Readings on ANY Table R1 Effluent Monitor> Table R1 value

1. Reading on the Liquid Radwaste Effluent Monitor (17RM-350) > 2 times hi-hi trip for? 60 minutes.

OR

.!:! 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes. 2. Readings on ANY Table R1 Effluent Monitor> Table R1 value Cl for~ 15 minutes. for~ 15 minutes. for~ 60 minutes.

0 OR OR 0 OR 2.

  • Dose assessment using actual meteorology indicates doses at OR

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

2. Dose assessment using actual meteorology indicates doses at 3. Confirmed sample analyses for gaseous or liquid releases ra or beyond the site boundary of EITHER: a. > 1 O mRem TEDE 0:: or beyond the site boundary of EITHER: indicate concentrations or release rates
a. > 1000 mRem TEDE OR
a. > 100 mRem TEDE > 2 times ODCM Limit with a release duration OR b. > 50 mRem CDE Thyroid of 2:, 60 minutes.
b. > 5000 mRem CDE Thyroid OR OR OR b. > 500 mRem CDE Thyroid 3. Analysis of a liquid effluent sample indicates a concentration
3. Field survey results at or beyond the site boundary indicate or release rate that would result in doses greater than OR EITHER of the following at or beyond the site boundary.

EITHER:

a. Gamma (closed window) dose rates 3. Field survey results at or beyond the site boundary indicate a. 1 O mRem TEDE for 60 minutes of exposure.

EITHER: OR

> 1000 mR/hr are expected to continue

b. 50 mRem CDE Thyroid for 60 minutes for::. 60 minutes. a. Gamma (closed window) dose rates of exposure.

OR > 100 mR/hr are expected to continue for~ 60 minutes. OR

b. Analyses of field survey samples indicate 4. Field survey results at or beyond the site boundary indicate

> 5000 mRem CDE Thyroid for 60 minutes OR EITHER:

of inhalation. a. Gamma (closed window) dose rates

b. Analyses of field survey samples indicate

> 500 mRem CDE Thyroid for 60 minutes > 1O mR/hr are expected to continue of inhalation. for 2:, 60 minutes.

OR

b. Analyses of field survey samples indicate

> 50 mRem CDE Thyroid for 60 minutes of inhalation.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-1 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex HOT MATRIX

  • HOT MATRIX
  • Exelon Nuclear Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency Site Area Emeraencv Alert Unusual Event Stack 7880 mR/hr 788 mR/hr 78.8 mR/hr 0.451 mR/hr (High Range Monitor)

Rx Bldg Exh N/A N/A N/A 9.50E+05 cpm (Low Range Monitor)

Turb Bldg Exh 2.44 mR/hr 0.244 mR/hr N/A 6.72E+05 cpm (Low Range Monitor)

Radw Bldg Exh 4.74 mR/hr 0.474 mR/hr N/A NIA Refuel Floor Exh N/A N/A N/A 9.28E+05 cpm (Low Range Monitor)

July 2019 HOT MATRIX JAF 2-2 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Abnormal Rad Levels/ Radiological Effluents RG2 RS2 RA2 RU2 Spent fuel pool level cannot be restored to at least 1.00 foot Spent fuel pool level at 1.00 foot. Significant lowering of water level above, or damage to, UNPLANNED loss of water level above irradiated fuel.

for 60 minutes or longer. irradiated fuel.

Emergency Action Level (EAL): Emergency Action Level (EAL):

Emergency Action Levels (EALl: Emergency Action Level {EALl:

Lowering of spent fuel pool level to 1.00 foot as indicated

1. a. UNPLANNED water level drop in the REFUELING Note: The Emergency Director should declare the event on 19Ll-60A or 19Ll-60B.

PATHWAY as indicated by ANY of the following:

promptly upon determining that the applicable time 1. Uncovery of irradiated fuel in the REFUELING has been exceeded. or will likely be exceeded. PATHWAY.

  • Inability to restore and maintain Spent Fuel Pool water level > low water level alarm.

OR Spent fuel pool level cannot be restored to at least OR 1.00 foot as indicated on19Ll-60A or 19Ll-60B 2. Damage to irradiated fuel resulting in a release of for~ 60 minutes. radioactivity from the fuel as indicated by ANY

  • Indication or report of a drop in water level Table R2 Refuel Floor Radiation Monitors Table R2 Radiation Monitor Alarm. in the REFUELING PATHWAY.

OR AND

3. Lowering of spent fuel pool level to 11.00 feet as b. UNPLANNED Area Radiation Monitor reading rise

J!l C:

I
i:

w RA3 RU3

<a Radiation levels that impede access to equipment necessary Reactor coolant activity greater than Technical Cl for normal plant operations, cooldown or shutdown. Specification allowable limits.

0 Table R4

§ Areas with Entry Related Mode Applicability Emergency Action Level {EAL): Emergency Action Level (EAL):

Table R3 "C

Areas Requiring Continuous Occupancy Entry Note: If the equipment in the room or area listed in Table

&. Related R4 was already inoperable, or out of service, before 1 Offgas radiation?. hi-hi alarm

  • Main Control Room - (by survey) Area the event occurred, then no emergency classification Mode
  • Central Alarm Station - (by survey) Applicability is warranted. OR
1. Dose rate> 15 mR/hr in ANY of the areas in Table R3. 2. Specific coolant activity> 2.0 µCi/gm 1-131 dose
  • Reactor Building East Crescent equivalent.

OR

  • Reactor Building West Crescent
  • Reactor Building 272' Elevation 2. UNPLANNED event results in radiation levels that Mode 3, 4, prohibit or significantly impede access to ANY of the
  • Reactor Building 300' Elevation and 5 areas in Table R4.
  • Relay Room
  • North Cable Room Modes: 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuehng D-Defueled July 2019 HOT MATRIX JAF 2-3 HOT MATRIX EP-AA-1014Addendum 3 (Rev.1)

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity> 300 uCifgm 1-131 dose None None None

1. RCS Activity None None equivalent.
2. RPV water level cannot be restored and 1. RPV water lever cannot be restored and
1. SAOG entry required maintained> o inches (TAF). maintained > o inches (TAF). SAOG entry required
2. RPV Water Level None None OR OR
3. RPV water level .E!.!l!!.Q! be determined. 2. RPV water level £!!!!!.Q! be determined.
3. Primary Containment pressure> 56 psig.

OR

1. a. Primary Containment pressure 1. UNPLANNED rapid drop in Primary 4. a. Primary Containment hydrogen
3. Primary > 2.7 psig. Containment pressure following Primary concentration~ 6%.

Containment AND Containment pressure rise. AND None None None Pressure/ b. Primary Containment pressure rise is OR b. Primary Containment oxygen concentration Conditions due to RCS leak.age. 2. Primary Containment pressure response not ~5%.

consistent with LOCA conditions. OR

5. Heat Capacity Temperature Limit (HCTL)

(EOP-11) exceeded.

3. UNISOLABLE primary system leakage that results in EITHER of the following:
1. UNISOLABLE Main Steam line (MSL),

HPCI, RWCU, RCIC, or Feedwater line a. Secondary Containment area temperature break. > EOP-5 Maximum Normal Operating

4. RCS Leak Rate None None Limit. None None OR
2. Emergency RPV Depressurization is OR required. b. Secondary Containment area radiation

> EOP-S Maximum Normal Operating Limit.

5. Primary Drywell radiation monitor reading Drywell radiation monitor reading Drywell radiation monitor reading Containment None None None

> 1.8E+03 R/hr (1800 Rlhr). > 63 R/hr. > 1.8E+04 R/hr (18,000 R/hr).

Radiation

1. UNJSOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal.

OR

2. Intentional Primary Containment venting or purging per EOPs or SAOGs due to accident conditions.

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

a. Secondary Containment area temperature

> EOP-5 Maximum Safe Operating Limit.

OR

b. Secondary Containment area radiation

> EOP-5 Maximum Safe Operating Limit.

7. Emergency 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the Director Emergency Director that indicates Loss Emergency Director that indicates Emergency Director that indicates Loss of Emergency Director that indicates Potential Emergency Director that indicates Loss of the Emergency Director that indicates Potential Judgment of the Fuel Clad Barrier. Potential Loss of the Fuel Clad Barrier. the RCS Barrier. Loss of the RCS Barrier. Containment Barrier. Loss of the Containment Barrier.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D- Defueled July 2019 HOT MATRIX JAF 2-4 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear I

System Malfunction MG1 ij][][] MS1 ij][][] MA1 ij][][] MU1 ij][][]

Prolonged loss of all offsite and all onsite AC power to Loss of all offsite and onsite AC power to emergency Loss of all but one AC power source to emergency buses for Loss of all offsite AC power capability to emergency buses emergency buses. buses for 15 minutes or longer. 15 minutes or longer. for 15 minutes or longer.

Emergenc:t Action Level {EAL}: Emergenc:t Action Level {EAL}: Emergenc:t Action Level {EAL}: Emergenc:t Action Level {EAL}:

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

1. Loss of ALL offsite and onsite AC power to 4160 V 0

== 1. Loss of ALL offsite and onsite AC power to 4160 V 1. AC power capability to 4160 V emergency buses 10500 Loss of ALL offsite AC power capability 4160 V emergency

a. emergency buses 10500 and 10600.

and 10600 reduced to only one of the following power u

<(

2.

AND EITHER of the following:

emergency buses 10500 and 10600.

AND ..

sources for 2: 15 minutes.

Reserve Station Transfonner T-2 buses 10500 and 10600 for 2: 15 minutes.

Reserve Station Transformer T-2 0

Cl)

Cl) 0

a. Restoration of at least one 4160 V emergency bus 10500 or 10600 in< 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.
2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 in< 15 minutes from . Reserve Station Transformer T-3 Station Service Transfonner T-4 (While

. Reserve Station TransfonnerT-3 Station Service Transformer T-4 (While backfeeding from Main Transfonner)

...J OR the time of loss of both offsite and onsite AC power. backfeeding from Main Transfonner)

b. RPV water level cannot be restored and maintained EDGA

> -19 inches (MSCRWL).

.. EOG B EOG C AND

. EOG D Main Generator via T-4

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

MG2 IJJI]~ MS2 IJ][][]

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

longer.

Emergenc:t Action Level {EAL}:

Emergenc:t Action Level {EAL}:

Note: The Emergency Director should declare the event Note: The Emergency Director should declare the event promptly upon determining that the applicable time

>> promptly upon determining that the applicable time has has been exceeded, or will likely be exceeded.

0 been exceeded, or will likely be exceeded.

a.==

1. Loss of ALL offsite and onsite AC power to 4160 V Voltage is < 105 VDC on Vital DC buses 71 BCB-2A u and 71BCB-2B for;:: 15 minutes.

0 0

emergency buses 10500 and 10600.

AND Cl) 2. Voltage is < 105 VDC on Vital DC buses 71 BCB-2A and Cl) 0 71BCB-2B.

...J AND

3. All AC and Vital DC power sources in EALs #1 and #2 have been lost for;:: 15 minutes.

Modes. 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D- Defueled July 2019 HOT MATRIX JAF 2-5 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear System Malfunction MS3 MA3 III[] MU3 Inability to shutdown the reactor causing a challenge to RPV Automatic or manual scram fails to shutdown the Automatic or manual scram fails to shutdown the reactor.

water level or RCS heat removal. reactor, and subsequent manual actions taken at the Emergency Action Level (EAU:

Reactor Control Console are not successful in shutting Emergency Action Level (EAL):

down the reactor. Note: A manual action is any operator action, or set of

1. Automatic scram did not shutdown the reactor as actions, which causes the control rods to be rapidly Emergency Action Level (EAL!:

indicated by Reactor Power c:_ 2.5%. inserted into the core. This action does not include Note: A manual action is any operator action, or set of manually driving in control rods or implementation of AND boron injection strategies.

actions, which causes the control rods to be rapidly

2. ALL manual / ARI actions to shutdown the reactor have inserted into the core. This action does not include 1 a. Automatic scram did not shutdown the reactor as been unsuccessful as indicated by Reactor Power manually driving in control rods or implementation of
  • indicated by Reactor Power c:_ 2.5%.

c:_2.5%. boron injection strategies.

~ AND

.2 AND 1. Automatic or manual scram did not shutdown the

b. Subsequent manual/ ARI action taken at the "iii reactor as indicated by Reactor Power c:_ 2.5%. Reactor Control Console is successful in shutting
11. 3. EITHER of the following conditions exist:

Cl) AND down the reactor as indicated by Q.

  • RPV water level cannot be restored and maintained Reactor Power< 2.5%.

a: > -19 inches (MSCRWL). 2. Manual / ARI actions taken at the Reactor Control Console are not successful in shutting down the reactor OR OR as indicated by Reactor Power c:_ 2.5%. 2. a. Manual scram did not shutdown the reactor as

  • Heat Capacity Temperature Limit (HCTL) indicated by Reactor Power c:_ 2.5%.

(EOP-11) exceeded. AND

b. EITHER of the following:
1. Subsequent manual/ ARI action taken at the Reactor Control Console is successful in shutting down the reactor as indicated by Reactor Power< 2.5%.

OR

2. Subsequent automatic scram I ARI is successful in shutting down the reactor as indicated by Reactor Power< 2.5%.

MA4 MU4 II) UNPLANNED loss of Control Room indications for 15 UNPLANNED loss of Control Room indications for 15 C minutes or longer.

minutes or longer with a significant transient in progress.

~ltl Table M2 Significant Transients Emergency Action Level (EAL!: Emergency Action Level (EAL):

Table M1 Control Room Parameters

=c"

-=E

.. Reactor Power

  • Auto/Manual runback > 25% thermal reactor Note: The Emergency Director should declare the event promptly upon determining that the applicable time Note: The Emergency Director should declare the event promptly upon determining that the applicable time 0

0 a:

.. RPV Water Level RPV Pressure

  • power Electric load rejection > 25% full electric has been exceeded, or will likely be exceeded. has been exceeded, or will likely be exceeded.

e.. . Primary Containment Pressure load 1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room u

C 0

. Torus Level Torus Temperature Reactor Scram ECCS injection for c:_ 15 minutes .

AND for c:_ 15 minutes.

  • Thermal Power oscillations > 10% (peak to 2. ANY Table M2 transient in progress.

peak)

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling 0- Defueled July 2019 HOT MATRIX JAF 2-6 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNU$UAL EVENT
  • Exelon Nuclear System Malfunction MA5 [j]~~

Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.

Emergency Action Level (EAL):

Note:

. This EAL is only applicable to SAFETY SYSTEMs

. having two (2) or more trains.

If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred,

. then this emergency classification is not warranted.

If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not

. warranted.

If a hazardous event occurs and it is determined that the conditions of MA5 are not met, then assess the event via HU3, HU4, or HUB.

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

Seismic event (earthquake)

.. Internal or external flooding event High winds or tornado strike

. . FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager AND

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

E

,S! AND 1/)

en i:,

J!!

C.

EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the ca SAFETY SYSTEM required by Technical en Specifications for the current operating mode.

J!l tJ OR

~ca .

'C ca N

Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the ca current operating mode.

i::

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 HOT MATRIX JAF 2-7 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear System Malfunction MUG BJ~[]

RCS leakage for 15 minutes or longer.

Emergenc:t Action Level {EAL}:

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

1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for:::_ 15 minutes.

OR

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

OR

3. Leakage from the RCS to a location outside the Drywell > 25 gpm for:::_ 15 minutes.

MU7 BJ~[]

Table M3 Communications Capability Loss of all onsite or offsite communication capabilities.

System Onsite Offsite NRC Emergenc:t Action Level {EAL}:

Page/Party System X 1. Loss of ALL Table M3 onsite communication (Gaitronics) capabilities affecting the ability to perform routine Control Room/Portable X operations.

Radio OR

2. Loss of ALL Table M3 offsite communication Plant Telephones (all X X X capabilities affecting the ability to perform offsite C

.!2 VOiP, switched, non-switched) notifications.

OR "lii u Installed Out-of-Plant X X X 3. Loss of ALL Table M3 NRG communication

  • c:l Cellular Phones capabilities affecting the ability to perform NRG E Plant Satellite Phones X X notifications.

E (Installed in CR and 0

(.) deployable)

RECS X Dedicated Phone Lines X X

(ENS)

HPN and FTS 2001 X X

Modes. 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019 HOT MATRIX JAF 2-8 HOT MATRIX EP-AA-1014Addendum 3 (Rev.1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and other conditions Affecting Plant Safety HS1 [j][?]@l@l[fil[g HA1 [j][?]@]@][fil[§ HU1 [j][?]@]@] [fil [§ HOSTILE ACTION within the PROTECTED AREA. HOSTILE ACTION within the OWNER CONTROLLED Confirmed SECURITY CONDITION or threat.

AREA or airborne attack threat within 30 minutes.

Emergencll: Action Level {EAL}: Emergencll: Action Level {EAL}:

Emergencll: Action Level {EAL}:

A notification from the Security Supervisor that a HOSTILE 1. Notification of a credible security threat directed at the C ACTION is occurring or has occurred within the 1. A validated notification from NRC from an aircraft site as detennined per SY-AA-101-132, Security

~.., PROTECTED AREA. attack threat< 30 minutes of the site . Assessment and Response to Unusual Activities.

~ OR OR

2. A validated notification from the NRC providing

~u, 2. Notification by the Security Supervisor that a infonnation of an aircraft threat.

HOSTILE ACTION is occurring or has occurred 0 OR

c within the OWNER CONTROLLED AREA.
3. Notification by the Security Supervisor of a SECURITY CONDITION that does not involve a HOSTILE ACTION.

HS2 [j] [?]@]@] [fil [§ HA2 [j] [?] §@I [fil [§ Inability to control a key safety function from outside the Control Room evacuation resulting in transfer of plant control Table H1 Safety Functions Control Room. to alternate locations.

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

Emergencll: Action Level {EAL}: Emergencll: Action Level {EAL}:

0

.l:; . RPV Water Level (ability to cool the core)

Note: The Emergency Director should declare the event promptly upon determining that the applicable time has A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the C

0 been exceeded, or will likely be exceeded.

u RCS Heat Removal (ability to maintain heat sink) Control Room.

C

~ 1. A Control Room evacuation has resulted in plant control

...a: being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the

...0 Control Room .

.S!u, C AND

! 2. Control of ANY Table H1 key safety function is not I-reestablished in < 30 minutes.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-9 HOT MATRIX EP-AA-1014Addendum 3 (Rev.1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HU3 Table H2 Areas

. Reactor Building (when inerted the FIRE potentially degrading the level of safety of the plant.

Emergency Action Level (EAL):

. Drywell is exempt)

Control Room / Relay Room I Cable Run Rooms I Cable Spreading Note:

  • The Emergency Director should declare the

.. Room Electric Bays event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

.. Control Room AC Equipment Room Control Room Chiller Room Emergency Diesel Generator

  • Escalation of the emergency classification level would be via IC CA2 or MA5.

. Building Battery Rooms / Battery Room

1. A FIRE in ANY Table H2 area is not extinguished in < 15 minutes of ANY of the following FIRE detection Corridor indications:

RHRSW / ESW Pump Rooms

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

Cable Tunnels

. Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MSIV /

Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm ADS) OR

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

AND

b. The existence of a FIRE is not verified in< 30 minutes of alarm receipt.

OR

3. A FIRE within the plant PROTECTED AREA not extinguished in < 60 minutes of the initial report, alarm or indication.

OR

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

Modes. 1 - Power Operatron 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-10 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HU4 IT] ill[}]@]@] [g Seismic event greater than QBE levels.

Emergencl£ Action Level {EAL}:

Note:

. For emergency classification if EAL# 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Emergency Director in!:. 15 minutes of the event.

. Escalation of the emergency classification level would be via IC CA2 or MA5.

1. Seismic event> Operating Basis Earthquake (OBE) as determined by seismic monitoring system in accordance with AOP-14 Earthquake.

OR

2. When Seismic Monitoring Equipment is not available:
a. Control Room personnel feel an actual or potential (I) seismic event.

ra AND

l Cl"

.c: b. ANY one of the following confirmed in!:. 15 minutes of t:: the event:

w ra

. 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 a manner deemed appropriate by the Shift Manager or Emergency Director.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-11 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety Table H3 HAS lfil@Jrn Areas with Entry Related Mode Applicability Gaseous release impeding access to equipment Entry Related necessary for normal plant operations, cooldown or Area Mode shutdown.

u, Annlicability Emergency Action Level (EALl:

C)

"0

'i(

.. Reactor Building East Crescent Note: If the equipment in the listed room or area was already inoperable, or out of service, before the Reactor Building West Crescent I- event occurred, then no emergency classification Reactor Building 272' Elevation is warranted.

Reactor Building 300' Elevation Mode 3, 4, and 5

. Relay Room North Cable Room

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

AND

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

HU6 II]~~@J~[l Hazardous Event Emergency Action Level (EAL}:

Note:

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

. Escalation of the emergency classification level would C

be via IC CA2 or MA5 .

Cl) 1. Tornado strike within the PROTECTED AREA.

w > OR u,

, 2. Internal room or area flooding of a magnitude sufficient to 0 require manual or automatic electrical isolation of a "Cl Ill SAFETY SYSTEM component required by Technical N Specifications for the current operating mode.

Ill J: OR

3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Intake Water Level > 255 feet.

OR

6. ESW intake bay water level ~ 237 feet.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019 HOT MATRIX JAF 2-12 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HG7 HS7 HU7 Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL Emergency Director warrant declaration of a SITE AREA Emergency Director warrant declaration of an ALERT. Emergency Director warrant declaration of an UNUSUAL EMERGENCY. EMERGENCY. EVENT.

Emergency Action Level {EAL):

Emergency Action Level (EAL): Emergency Action Level (EAL): Emergency Action Level {EAL):

Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director, indicate that events are in progress or Other conditions exist which in the judgment of the Emergency Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve an actual or potential Director indicate that events are in progress or have occurred have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of substantial degradation of the level of safety of the plant or a which indicate a potential degradation of the level of safety of core degradation or melting with potential for loss of plant functions needed for protection of the public or HOSTILE security event that involves probable life threatening risk to the plant or indicate a security threat to facility protection has containment integrity or HOSTILE ACTION that results in an ACTION that results in intentional damage or malicious acts, site personnel or damage to site equipment because of been initiated. No releases of radioactive material requiring actual toss of physical control of the facility. Releases can be (1) toward site personnel or equipment that could lead to the HOSTILE ACTION. Any releases are expected to be limited offsite response or monitoring are expected unless further reasonably expected to exceed EPA Protective Action likely failure of or, (2) that prevent effective access to to small fractions of the EPA Protective Action Guideline degradation of safety systems occurs.

Guideline exposure levels offsite for more than the immediate equipment needed for the protection of the public. Any exposure levels.

site area. releases are not expected to result in exposure levels which

....C exceed EPA Protective Action Guideline exposure levels (II beyond the site boundary.

E Cl "C

~

e c

~

C (II e>

(II E

w Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuehng D-Defueled July 2019 HOT MATRIX JAF 2-13 HOT MATRIX EP-AA-1014Addendum 3 (Rev.1)

James A. FitzPatrick Nuclear Power Plant Annex HOT MATRIX

  • HOT MATRIX
  • Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 l!]~@l@l@][g Damage to a loaded cask CONFINEMENT BOUNDARY.

Emergencl£ Action Level {EAL}:

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading> ANY Table E-1 values:

Table E-1 Radiation Reading Overpack Overpack Overpack Overpack Serial Average Serial Average Surface Number Surface Dose Number Dose Rates Rates mrem/hr mrem/hr (gamma+neutron) HI- (gamma+neutron)

STORM 100S (XXX)

HI- 80 on the side SIN- 220 on the side STORM 0186, 20 on the top 40 on the top 100S 0187, en LL 32 at the inlet 0188 SIN-15,

!!? and outlet 16, 17 vent ducts SIN-0307, 0308, 0309, 0310, 0311, HI- 100 on the side 0312, STORM 20 on the top 0679, 100S 600 on the side 0680, (232) 90 at the inlet 0681, 60 on the top SIN- and outlet 0682, vent ducts 0683, 0169, 0170, 0690, 0171 0691, 0692, 0693, 0694, 0695 Modes. 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019 HOT MATRIX JAF 2-14 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels/ Radiological Effluents RG1 [j][g]@l@J[§J[g RS1 [j][g]@l@J[§J [g RA1 III [g]@l@J[§J ITl RU1 [j][g]@l@J[§J [g Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite dose Release of gaseous or liquid radioactivity greater than 2 than 1000 mRem TEOE or 5000 mRem thyroid COE. than 100 mRem TEOE or 500 mRem thyroid COE. greater than 10 mRem TEOE or 50 mRem thyroid COE. times the OOCM limits for 60 minutes or longer.

Emergency Action Level (EAL): Emergency Action Level (EAL): Emergency Action Level (EAL): Emergency Action Level (EAL}:

Notes:

The Emergency Director should declare the event promptly upon detenmining that the applicable time has been Notes:

. The Emergency Director should declare the event promptly upon detenmining that the applicable time has been Notes:

The Emergency Director should declare the event promptly upon detenmining that the applicable time has been exceeded.

Notes:

The Emergency Director should declare the event promptly upon detenmining that the applicable time has

. 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, or will likely be exceeded.

If an ongoing release is detected and the release start time is unknown, assume that the release duration has

. or will likely be exceeded.

lf an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15

. 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

. exceeded 15 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the

. minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent

. exceeded 60 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is effluent flow past an effluent monitor is known to have effluent flow past an effluent monitor is known to have flow past an effluent monitor is known to have stopped due to established. If the effluent flow past an effluent monitor stopped due to actions to isolate the release path, then the stopped due to actions to isolate the release path, then the actions to isolate the release path, then the effluent monitor is known to have stopped due to actions to isolate the

.l!l C

(I) effluent monitor reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in .

effluent monitor reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in

. reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in EAL

  1. 1 (Table R1) should be used for emergency classification release path, then the effluent monitor reading is no longer valid for classification purposes.

EAL #1(Table R1) should be used for emergency EAL #1 (Table R 1) should be used for emergency assessments until the results from a dose assessment using

=

I classification assessments until the results from a dose classification assessments until the results from a dose actual meteorology are available. 1. Reading on the Liquid Radwaste Effluent Monitor w assessment using actual meteorology are available. assessment using actual meteorology are available. (17RM-350) > 211mes hi-hi trip for le 60 minutes.

iii 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for OR 0 .::. 15 minutes.

  • a, 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value 2. Readings on ANY Table R1 Effluent Monitor> Table R1 0 for 2:. 15 minutes. for.::, 15 minutes. OR value for 2:. 60 minutes.

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

'ti 2. Dose assessment using actual meteorology indicates doses at

2. Dose assessment using actual meteorology indicates doses at ca or beyond the site boundary of EITHER: a. > 10 mRem TEOE 3. Confirmed sample analyses for gaseous or liquid releases D: or beyond the site boundary of EITHER: OR indicate concentrations or release rates
a. > 1000 mRem TEDE
a. > 100 mRem TEDE b. > 50 mRem CDE Thyroid > 2 times ODCM Limit with a release duration OR OR of? 60 minutes.
b. > 5000 mRem CDE Thyroid OR
3. Analysis of a liquid effluent sample indicates a concentration or OR b. > 500 mRem CDE Thyroid release rate that would result in doses greater than EITHER of
3. Field survey results at or beyond the site boundary indicate the following at or beyond the site boundary OR EITHER: a. 1o mRem TEDE for 60 minutes of exposure
a. Gamma (closed window) dose rates 3. Field survey results at or beyond the site boundary indicate EITHER: OR

> 1000 mR/hr are expected to continue b. 50 mRem CDE Thyroid for 60 minutes for,:: 60 minutes. a. Gamma (closed window) dose rates of exposure OR > 100 mR/hr are expected to continue OR

b. Analyses of field survey samples indicate for.::. 60 minutes. 4. Field survey results at or beyond the site boundary indicate

> 5000 mRem CDE Thyroid for 60 minutes OR EITHER:

of inhalation. a. Gamma (closed window) dose rates

b. Analyses of field survey samples indicate

> 500 mRem CDE Thyroid for 60 minutes > 1 O mR/hr are expected to continue of inhalation. for.::, 60 minutes.

OR

b. Analyses of field survey samples indicate

> 50 mRem CDE Thyroid for 60 minutes of inhalation.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling 0-0efueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-15 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency Site Area Emergency Alert Unusual Event Stack 7880 mR/hr 788 mR/hr 78.8 mR/hr 0.451 mR/hr (High Range Monitor)

Rx Bldg Exh N/A NIA N/A 9.50E+05 cpm (Low Range Monitor)

Turb Bldg Exh 2.44 mR/hr 0.244 mR/hr N/A 6.72E+05 cpm (Low Range Monitor)

Radw Bldg Exh 4.74 mR/hr 0.474 mR/hr N/A N/A Refuel Floor Exh N/A N/A N/A 9.28E+05 cpm (Low Range Monitor)

July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-16 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear I

Abnormal Rad Levels/ Radiological Effluents RG2 RS2 RA2 RU2 Spent fuel pool level cannot be restored to at least 1.00 foot Spent fuel pool level at 1.00 foot. Significant lowering of water level above, or damage to, UNPLANNED loss of water level above irradiated fuel.

for 60 minules or longer. irradiated fuel.

Emergency Action Level (EAL): Emergency Action Level (EAL):

Emergency Action Levels (EAU: Emergency Action Level (EAU:

Lowering of spent fuel pool level to 1.00 foot as indicated 1. a. UNPLANNED water level drop in the REFUELING Note: The Emergency Director should declare the event on 19Ll-60-A or 19Ll-60B. PATHWAY as indicated by ANY of the following:

promptly upon determining that the applicable time 1. Uncovery of irradiated fuel in the REFUELING has been exceeded, or will likely be exceeded. PATHWAY.

  • Inability to restore and maintain Spent Fuel Pool water level> low water level alarm.

OR Spent fuel pool level cannot be restored to at least

2. Damage to irradiated fuel resulting in a release of OR 1.00 foot as indicated on 19Ll-60A or 19Ll-60B for~ 60 minutes. radioactivity from the fuel as indicated by ANY
  • Indication or report of a drop in water level in Table R2 Refuel Floor Radiation Monitors Table R2 Radiation Monitor Alarm. the REFUELING PATHWAY.

OR AND

3. Lowering of spent fuel pool level to 11.00 feet as b. UNPLANNED Area Radiation Monitor reading rise
  • 17RIS-456A or B Refuel Floor Exhaust RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

Table R4 Emergency Action Level (EAU:

Areas with Entry Related Mode Applicability Table R3 Note: If the equipment in the room or area listed in Table Areas Requiring Continuous Occupancy Entry R4 was already inoperable, or out of service, before Related the event occurred, then no emergency classification

  • Main Control Room - (by survey) Area Mode is warranted.
  • Central Alarm Station - (by survey) Applicability
1. Dose rate> 15 mR/hr in ANY of the areas in Table R3.
  • Reactor Building East Crescent OR
  • Reactor Building West Crescent 2. UNPLANNED event results in radiation levels that
  • Reactor Building 272' Elevation prohibit or significantly impede access to ANY of the Mode 3, 4, areas in Table R4.
  • Reactor Building 300' Elevation and 5
  • Relay Room
  • North Cable Room Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-17 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)
Ja:m:=e::s::A::*::::F::itz:P:::::::at::r"::1c:::k:::N::u::c:le:a::r:::P:::o:::w::e::r:::P:::l:an=t=A=n=n=e=x======C=O=L=D=S=H=U='T~D=O~W':'::N':'/R~E=FU~E=L=IN::G~M=A=T='R~l=X'========================--'C"-O~L.,D:...,S,,,Hc.Us,.T-'-"'D-"O'-"'WN/REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Exelon Nuclear Cold Shutdown / Refueling System Malfunctions CA1 ~@Jig CU1 ~@Jig Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses for for 15 minutes or longer. 15 minutes or longer.

EmergencJt: Action Level {EAL}: EmergencJt: Action Level {EAL}:

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

-- 1. Loss of ALL offsite and onsite AC power to 4160 V 1. AC power capability to 4160 V emergency buses 10500 and emergency buses 10500 and 10600. 10600 reduced to only one of the following power sources AND for::_ 15 minutes.

Reserve Station Transformer T-2

2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 in< 15 minutes from the time of loss of both offsite and onsite AC power.

. Reserve Station Transformer T-3 Station Service Transformer T-4 (While backfeeding

.. from Main Transformer)

EDGA

...a, .. EDG B EDGC EDGD 0 AND D..

2. ANY additional single power source failure will result in a 0

cl:

0 Cl)

Cl) 0 loss of ALL AC power to SAFETY SYSTEMS.

..J Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuellng D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-18 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear UNUSUAL EVENT Cold Shutdown/ Refueling System Malfunctions CA2 1+/-1~

Hazardous event affecting SAFETY SYSTEM required for the current operating mode.

Emergencl£ Action Level (EAL}: I Note:

. This EAL is only applicable to SAFETY SYSTEMs

. having two (2) or more trains.

If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred, then

. this emergency classification is not warranted.

If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not

. warranted.

If a hazardous event occurs and it is determined that the conditions of CA2 are not met, then assess the event via HU3, HU4, or HU6.

E 1. a. The occurrence of ANY of the following hazardous

.fl (I) events:

Seismic event (earthquake)

II)

Internal or external flooding event

.a, High winds or tornado strike

~ FIRE nl II)

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

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

AND C.

EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

OR

. Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-19 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum _3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Cold Shutdown / Refueling System Malfunctions

...a, CU3 mm Loss of Vital DC power for 15 minutes or longer.

==

0 Emergency Action Level (EAL}:

D.

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

Voltage is < 105 VDC on required Vital DC buses 71 BCB-2A and 71 BCB-2B for~ 15 minutes.

Table C1 Communication Capabilities CU4 m@ll9 Loss of all onsite or offsite communication capabilities.

Svstem Onsite Offsite NRC Page/Party System X Emergency Action Level (EAL}:

(Gaitronics) 1. Loss of ALL Table C1 onsite communication U)

C Control Room/Portable capabilities affecting the ability to perform routine X

~m Radio operations.

u Plant Telephones (all VOiP, X X X OR

  • c~

switched, non~switched)

2. Loss of ALL Table C1 offsite communication E Installed Out-of-Plant X X X capabilities affecting the ability to perform offsite E Cellular Phones notifications.

0 CJ Plant Satellite Phones X X OR (Installed in CR and deployable) 3. Loss of ALL Table C1 NRC communication capabilities affecting the ability to perform NRG RECS X notifications.

Dedicated Phone Lines X X (ENS)

HPN and FTS 2001 X X Table C2 RCS Heat-up Duration Thresholds CA5 mm CU5 mm Inability to maintain the plant in cold shutdown. UNPLANNED rise in RCS temperature.

RCS Containment Heat-up Emergency Action Levels (EAL): Emergency Action Levels (EAU:

Status Closure Status Duration Intact Not APolicable 60 minutes* Note:

. The Emergency Director should declare the event Note:

The Emergency Director should declare the event promptly upon determining that the applicable time promptly upon determining that the applicable time

.>i:

in C

Not Intact Established 20 minutes*

has been exceeded, or will likely be exceeded. . has been exceeded, or will likely be exceeded.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature m A momentary UNPLANNED excursion above the a, Technical Specification cold shutdown temperature limit when heat removal function is available does

c Not Established 0 minutes limit when heat removal function is available does not warrant classification.

not warrant classification. 1. UNPLANNED rise in RCS temperature> 212 °F.

  • If an RCS heat removal system is in operation 1. UNPLANNED rise in RCS temperature> 212 °F OR within this time frame and RCS temperature is being reduced, then EAL #1 is not applicable.

for> Table C2 duration.

OR 2.

Loss of the following for~ 15 minutes.

ALL RCS temperature indications

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

. AND ALL RPV water level indications Modes. 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuellng D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-20 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

,.Ja,,,m=es"-"A,,_. .,_F.:.:ilz.,.P,._a,,,t,,_ri,,,c,,k_,_N,.,u,.,c:.::lee,a:,.r_,_P.ao,_,w,.,e,,.r_,_P-'la,,,n"'t'-'A"'n"'n"'e"'x'-----"C"'O'-'L::D=S.:.H:.,Uc.:T..eD,.,O~W=N"'/R.:.E::F,._U=E=L"-IN,.,G,,_,_,M.,A;,.Tc.:R..::l"'X'----------=-==~=========""'C""O"""'L"'D-'S"'H""U""T""D"""'O""W.,N/REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Exelon Nuclear Cold Shutdown / Refueling System Malfunctions CG6 1+/-1~ CS6 1+/-1~ CA6 1+/-1~ CU6 1+/-1~

Loss of RPV inventory affecting fuel clad integrity with Loss of RPV inventory affecting core decay heat removal Loss of RPV inventory. UNPLANNED loss of RP inventory for 15 minutes or containment challenged. capability. longer.

EmergencJl Action Level (EAL}:

EmergencJl Action Level (EAL}: EmergenCJl Action Level (EAL}: EmergenCJl Action Level (EAL}:

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

1. Loss of RPV inventory as indicated by 1 a. RPV water level < O inches (TAF) for 1. With CONTAINMENT CLOSURE not established, level< 126.5 inches. 1. UNPLANNED loss of reactor coolant results in the
30 minutes. RPV water level< 120.5 inches. inability to restore and maintain RPV level to above the OR AND OR procedurally established lower limit
2. With CONTAINMENT CLOSURE established, RPV 2. a. RPV water level Efil!lli!! be monitored for::: 15 minutes.
b. ANY Table C4 Containment Challenge Indication.

water level< 0 inches (TAF). for::: 15 minutes.

OR OR OR AND

2. a. RPV water level cannot be monitored 3. a. RPV water level cannot be monitored for 2. a. RPV water level cannot be monitored.

for::: 30 minutes-.- - b. Loss of RPV inventory per Table C3 indications.

30 minutes. AND

~ AND AND

b. Loss of RPV inventory per Table C3 indications.

.s b. Core uncovery is indicated by ANY of the C

GI

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

Table C3 indication of a sufficient magnitude following:

- to indicate core uncovery. Table C3 indication of a sufficient magnitude

.!: to indicate core uncovery.

GI CD I'll . OR 18RIA-052-30 Refuel Floor West (EPIC A-1247) Rad monitor::: 3 R/hr.

. OR 1BRIA-052-30 Refuel Floor West (EPIC A-I'll 1247) Rad monitor::: 3 R/hr.

GI

..J AND U) c. ANY Table C4 Containment Challenge Indication.

0 a::

. Table C3 Indications of RCS Leakage UNPLANNED Drywell equipment drain sump level Table C4 Containment Challenge Indications rise*

UNPLANNED Drywell floor drain sump level rise*

. Primary Containment Hydrogen Concentration

. UNPLANNED Reactor Building equipment sump level rise* .

6% and Oxygen ::: 5%

UNPLANNED rise in containment pressure

. UNPLANNED Reactor Building floor drain sump level * . CONTAINMENT CLOSURE.!!.!!! established*

rise*

UNPLANNED Torus level rise*

. Secondary Containment area radiation > ANY Maximum Safe Operating Limit (EOP-5)

. UNPLANNED RPV make up rate rise*

. Observation of leakage or inventory loss

  • if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not
  • Rise in level is attributed to a loss of RPV inventory required.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuehng D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-21 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HS1 HA1 HU1 HOSTILE ACTION within the PROTECTED AREA. HOSTILE ACTION within the OWNER CONTROLLED Confirmed SECURITY CONDITION or threat.

AREA or airborne attack threat within 30 minutes.

Emergency Action Level {EAL): Emergency Action Level {EAL):

Emergency Action Level {EAL):

A notification from the Security Supervisor that a HOSTILE 1. Notification of a credible security threat directed at ACTION is occurring or has occurred within the 1. A validated notification from NRC of an aircraft the site as determined per SY-AA-101-132. Security PROTECTED AREA. attack threat< 30 minutes from the site. Assessment and Response to Unusual Activities.

OR OR

2. A validated notification from the NRC providing
2. Notification by the Security Supervisor that a information of an aircraft threat.

HOSTILE ACTION is occurring or has occurred OR within the OWNER CONTROLLED AREA.

3. Notification by the Security Supervisor of a SECURITY CONDITION that does !!Q! involve a HOSTILE ACTION.

HS2 HA2 Inability to control a key safety function from outside the Control Room evacuation resulting in transfer of plant Table H1 Safety Functions Control Room. control to alternate locations.

Emergency Action Level {EAL): Emergency Action Level {EAL):

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

A Control Room evacuation has resulted in plant control

~C:

Note: The Emergency Director should declare the event

  • RPV Water Level (ability to cool the core) promptly upon determining that the applicable time being transferred from the Control Room to alternate 0 has been exceeded, or will likely be exceeded. locations per AOP-43, Plant Shutdown from Outside the u

....

  • RCS Heat Removal (ability to maintain heat sink) Control Room.

C:

ra 1. A Control Room evacuation has resulted in plant control ii: being transferred from the Control Room to alternate 0

~

. locations per AOP-43, Plant Shutdown from Outside the Control Room.

Ill C: AND E 2. Control of ANY Table H1 key safety function is not I-reestablished in < 30 minutes.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-22 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY Hazards and Other conditions Affecting Plant Safety HU3 Table H2 Areas Reactor Building (when inerted the Fl RE potentially degrading the level of safety of the plant.

. Drywell is exempt)

Control Room / Relay Room / Cable Run Rooms/ Cable Spreading Emergency Action Level (EAL):

Note:

Room

  • The Emergency Director should declare the

. event promptly upon determining that the

.. Electric Bays Control Room AC Equipment Room applicable time has been exceeded, or will likely be exceeded.

. Control Room Chiller Room Emergency Diesel Generator

  • Escalation of the emergency classification level would be via IC CA2 or MA5

. Building Battery Rooms / Battery Room 1. A FIRE in ANY Table H2 area is not extinguished in Corridor < 15 minutes of ANY of the following FIRE detection RHRSW / ESW Pump Rooms indications:

Cable Tunnels

  • Report from the field (i.e., visual observation)
  • Receipt of multiple (more than 1) fire alarms or Remote Safe Shutdown Panels indications
  • Field verification of a single fire alarm ADS)

OR

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

AND

b. The existence of a FIRE is not verified in

< 30 minutes of alarm receipt OR

3. A FIRE within the plant PROTECTED AREA not extinguished in< 60 minutes of the initial report, alarm or indication.

OR

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

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuelmg D - Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-23 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HU4 ITI~§@l@l[g Seismic event greater than OBE levels.

Emergencll Action Level (EAL!:

Note:

. For emergency classification if EAL# 2.b is not able to be confirmed. then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Emergency Director in:;_ 15 minutes of the event

. Escalation of the emergency classification level would be via IC CA2 or MA5.

1. Seismic event> Operating Basis Earthquake (OBE) as determined by seismic monitoring system in accordance with AOP-14 Earthquake.

OR

2. When Seismic Monitoring Equipment is not available:
a. Control Room personnel feel an actual or potential a, seismic event.

ra

, AND er

.c: b. ANY one of the following confirmed in:;_ 15 minutes t:: of the event:

ra w . 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 a manner deemed appropriate by the Shift Manager or Emergency Director.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-24 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HA5 @:]@][§ Table H3 Gaseous release impeding access to equipment Areas with Entry Related Mode Applicability necessary for normal plant operations. cooldown or shutdown.

Entry Ill Related Emergencll Action Level (EALj:

Ill Area

(!) Mode Note: If the equipment in the listed room or area was u Aoolicabilitv "i( already inoperable, or out of service, before the 0

I- .. Reactor Building East Crescent event occurred, then no emergency classification is warranted .

Reactor Building West Crescent Reactor Building 272' Elevation 1. Release of a toxic, corrosive, asphyxiant or Mode 3, 4, flammable gas in a Table H3 area.

Reactor Building 300' Elevation

. Relay Room North Cable Room and 5 2.

AND Entry into the room or area is prohibited or impeded HUS ITI~@l@llfil§ Hazardous Event Emergencll Action Level (EAL}:

Note:

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

Escalation of the emergency classification level would be via IC CA2 or MA5 C

a, w> 1. Tornado strike within the PROTECTED AREA.

Ill OR

s 2. Internal room or area flooding of a magnitude sufficient to 0

'E require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical

~ Specifications for the current operating mode.

Ill

c OR
3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Intake Water Level> 255 feet.

OR

6. ESW intake bay water level < 237 feet.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-25 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

  • GENERAL EMERGENCY I

SITE AREA EMERGENCY

.::J:::am=e=s=A=*::::F:::itz=P=at:::r::ic=k=N=u=c=le=a=r=P=o=w=e~r=P='l=an=t~A='::::n:::ne~x=====C~O=L;;!D~S=H=U~T~D,;,O~W;,,,;;N=/R~E,o,,F,""UcsE~L=IN=G~M~A=T"=R=l=X'========================="'C""'OcsL~D',,'S,a,HC!,'!U-',T~DcsO,_,,WN/REFUELING MATRIX ALERT UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HG7 HS7 HA7 HU7 Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL Emergency Director warrant declaration of a SITE AREA Emergency Director warrant declaration of an ALERT. Emergency Director warrant declaration of an UNUSUAL EMERGENCY. EMERGENCY. EVENT.

Emergency Action Level {EAL):

Emergency Action Level (EAU: Emergency Action Level (EAL): Emergency Action Level (EAL):

Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director, indicate that events are in progress or Other conditions exist which in the judgment of the Emergency Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve an actual or potential Director indicate that events are in progress or have occurred have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of substantial degradation of the level of safety of the plant or a which indicate a potential degradation of the level of safety of core degradation or melting with potential for loss of plant functions needed for protection of the public or HOSTILE security event that involves probable life threatening risk to the plant or indicate a security threat to facility protection has containment integrity or HOSTILE ACTION that results in an ACTION that results in intentional damage or malicious acts, site personnel or damage to site equipment because of been initiated. No releases of radioactive material requiring actual loss of physical control of the facility. Releases can be (1) toward site personnel or equipment that could lead to the HOSTILE ACTION. Any releases are expected to be limited offsite response or monitoring are expected unless further reasonably expected to exceed EPA Protective Action likely failure of or, (2) that prevent effective access to to small fractions of the EPA Protective Action Guideline degradation of safety systems occurs.

Guideline exposure levels offsite for more than the immediate equipment needed for the protection of the public. Any exposure levels.

site area. releases are not expected to result in exposure levels which

....C exceed EPA Protective Action Guideline exposure levels a, beyond the site boundary.

E Cl "t:I

~

f!

i5

~

C a,

I:'

a, E

w Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-26 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

.:J::a::m::::e::s::::A::.::::F::::itz=P=a=tr::ic::k::N::::u:::c:::l:::ea:::r::::P::::o:::w=e=r::::P::::l=a=nt==A=n=n=ex=======C=O=L=D=='S=H=U=T=D=O=W-=-'N=/=R=E=F=U=E=L"'IN.,_G"="'M"'A'"'T.,_R'"""IX=--==================~------"C-"O"'L"'D"'S"'H=U'-'T""D"'O'-'WN/REFUELING MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 ITI~@l@l@ll9 Damage to a loaded cask CONFINEMENT BOUNDARY.

Emergency Action Level (EAL):

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading> ANY Table E-1 values:

Table E-1 Radiation Reading Overpack Overpack Overpack Overpack Serial Average Serial Average Surface Number Surface Dose Number Dose Rates Rates mremlhr mrem/hr (gamma+neutron) HI- (gamma+neutron)

STORM 100S (XXX)

HI- 80 on the side SIN- 220 on the side STORM 0186.

100S 20 on the top 0187, 40 on the top iii u.. 32 at the inlet 0188 S/N-15,

!!? 16, 17 and outlet vent ducts SIN-0307, 0308, 0309, 0310, 0311, HI- 100 on the side 0312, STORM 20 on the top 0679, 100S 600 on the side 0680, (232) 90 at the inlet 0681, 60 on the top SIN- and outlet 0682, 0169, vent ducts 0683, 0170, 0690, 0171 0691, 0692, 0693, 0694, 0695 Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-27 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes.

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
a. > 1000 mRem TEDE OR
b. > 5000 mRem CDE Thyroid OR
3. Field survey results at or beyond the site boundary indicate EITHER:
a. Gamma (closed window) dose rates> 1000 mR/hr are expected to continue for~ 60 minutes.

OR

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

James A~ FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

~ RECOGNITION CATEGORY .

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG1 (cont)

Emergency Action Level (EAL) (cont):

Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency Stack 7880 mR/hr Turb Bldg Exh 2.44 mR/hr Radw Bldg Exh 4.74 mR/hr 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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. JAFNPP Technical Specifications Section 4. 1.1, Figure 4.4-1
3. OP-31 Process Radiation Monitoring Systems
4. DVP-01.02 Offsite Dose Calculation Manual
5. NEI 99-01 Rev 6, AG1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes.

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
a. > 100 mRem TEDE OR
b. > 500 mRem CDE Thyroid OR
3. Field survey results at or beyond the site boundary indicate EITHER:
a. Gamma (closed window) dose rates> 100 mR/hr are expected to continue for~ 60 minutes.

OR

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

July 2019 JAF 2-30 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS1 (cont)

Emergency Action Level (EAL) (cont):

Table R1 Effluent Monitor Thresholds Effluent Monitor Site Area Emergency Stack 788 mR/hr Turb Bldg Exh 0.244 mR/hr Radw Bldg Exh 0.474 mR/hr 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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RG1.

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. JAFNPP Technical Specifications Section 4.1.1, Figure 4.4-1
3. OP-31 Process Radiation Monitoring Systems
4. DVP-01.02 Offsite Dose Calculation Manual
5. NEI 99-01 Rev 6, AS1

.~ James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY-ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Exelon Nuclear RA1 Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mRem TEDE or 50 mRem thyroid COE.

Operating Mode Applicability:

1, 2, 3, 4, 5, D ,

Emergency Action Level (EAL):

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes.

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
a. > 10 mRem TEDE OR
b. > 50 mRem CDE Thyroid OR
3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
a. 10 mRem TEDE for 60 minutes of exposure OR
b. 50 mRem CDE Thyroid for 60 minutes of exposure

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY -

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 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.

OR

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

Table R1 Effluent Monitor Thresholds Effluent Monitor Alert Stack 78.8 mR/hr Turb Bldg Exh N/A

  • Radw Bldg Exh Basis:

N/A 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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RS1 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA1 (cont)

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. JAFNPP Technical Specifications Section 4.1.1, Figure 4.4-1
3. OP-31 Process Radiation Monitoring Systems
4. DVP-01.02 Offsite Dose Calculation Manual
5. NEI 99-01 Rev 6, AA1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • 1. Reading on the Liquid Radwaste Effluent Monitor (1 ?RM-350) > 2 times hi-hi trip for~ 60 minutes.

OR

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

Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Event Stack 0.451 mR/hr (High Range Monitor)

Rx Bldg Exh 9.50E+05 cpm (Low Range Monitor)

Turb Bldg Exh 6. 72E+05 cpm (Low Range Monitor)

Refuel Floor Exh 9.28E+05 cpm (Low Range Monitor)

OR

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY. -

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU1 (cont)

Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

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

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

The effluent monitor listed is normally used for planned discharges.

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 RA 1.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT RU1 {cont)

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. DVP-01.02 Offsite Dose Calculation Manual
3. OP-31 Process Radiation Monitoring Systems
4. NEI 99-01 Rev 6, AU1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG2 Initiating Condition:

Spent fuel pool level cannot be restored to at least 1.00 foot for 60 minutes or longer.

Operating Mode Applicability:

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

Note:

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

Spent fuel pool level cannot be restored to at least 1.00 foot as indicated on 19Ll-60A or 19Ll-60B 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. EOP-5 Secondary Containment Control
2. FSG-005, Alternate Spent Fuel Pool Makeup and Cooling
2. NEI 99-01 Rev 6, AG2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS2 Initiating Condition:

Spent fuel pool level at 1.00 foot Operating Mode Applicability:

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

Lowering of spent fuel pool level to 1.00 foot as indicated on 19Ll-60A or 19Ll-60B.

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. EOP-5 Secondary Containment Control
2. FSG-005, Alternate Spent Fuel Pool Makeup and Cooling
3. NEI 99-01 Rev 6, AS2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA2 Initiating Condition:

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

Operating Mode Applicability:

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

1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

OR

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R2 Radiation Monitor Alarm.

OR

3. Lowering of spent fuel pool level to 11.00 feet as indicated on 19Ll-60A or 19Ll-60B .

Table R2 Refuel Floor Radiation Monitors 18RIA-051-12 Spent Fuel Pool (EPIC A-1229) 18RIA-051-14 New Fuel Vault (EPIC A-1231)

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 or a significant lowering of water level within the spent fuel pool.

July 2019 JAF 2-40 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT RA2 (cont)

Basis (cont):

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.

EAL #1 Basis This EAL escalates from RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

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

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

EAL#2 Basis This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

EAL #3 Basis:

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

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

James A. FitzPafrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA2 (cont)

Basis Reference(s):

1. EOP-5 Secondary Containment Control
2. OP-32 Area Radiation Monitoring
3. JAFNPP EPG/SAG
4. FSG-005, Alternate Spent Fuel Pool Makeup and Cooling
5. NEI 99-01 Rev 6, AA2

James ,A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY*

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU2 Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.

Operating Mode Applicability:

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

1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
  • Inability to restore and maintain Spent Fuel Pool water level

> low water level alarm.

OR

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

AND

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

Table R2 Refuel Floor Radiation Monitors

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

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

This IC addresses a loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also July 2019 JAF 2-43 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION-CATEGORY-ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT RU2 (cont)

Basis (cont):

indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

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

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

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

  • Escalation of the emergency classification level would be via IC RA2.

Basis Reference(s):

1. AOP-53 Loss of Spent Fuel Pool, Reactor Cavity or Equipment Storage Pit Water Level
2. OP-32 Area Radiation Monitoring
3. OP-30 Fuel Pool Cooling and Cleanup System
4. ARP 09-3-1-9 Fuel Pool Cool & Cln Up Trouble
5. AOP-68 Spent Fuel Pool Trouble
6. NEI 99-01 Rev 6, AU2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

Table R3 Areas Requiring Continuous Occupancy

  • OR Main Control Room - (by survey)

Central Alarm Station - (by survey)

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the areas in Table R4.

Table R4 Areas with Entry Related Mode Applicability Entry Related Mode Area Applicability

  • Reactor Building East Crescent
  • Reactor Building West Crescent
  • Reactor Building 272' Elevation Mode 3, 4, and 5
  • Reactor Building 300' Elevation
  • Relay Room
  • North Cable Room

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY - -

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 (cont)

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal plant procedures. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

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

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

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

An emergency declaration is not warranted if any of the following conditions apply.

  • The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4 .

July 2019 JAF 2-46 EP-AA-1014 Addendum 3 (Rev. 1)

.- James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 (cont)

Basis (cont):

  • The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
  • The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
  • The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

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

Basis Reference(s):

1. JAFNPP Safe Shutdown Analysis
2. NEI 99-01 Rev 6, AA3

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU3 Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability:

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

1. Offgas radiation :::_ hi-hi alarm OR
2. Specific coolant activity> 2.0 µCi/gm 1-131 dose equivalent.

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.

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

  • Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.

Basis Reference(s):

1. DVP-01.02 Offsite Dose Calculation Manual Specification 3.6.1
2. Technical Specification 3.7.5
3. Technical Specification 3.4.6
4. Technical Specification Bases 3.4.6
5. OP-31 Process Radiation Monitoring
6. NEI 99-01 Rev 6, SU3

.~ James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear FG1 Initiating Condition:

Loss of ANY two barriers AND Loss or Potential Loss of the third barrier.

Operating Mode Applicability:

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

Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.

Basis:

Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the General Emergency classification level each barrier is weighted equally.

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

. RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:

Loss or Potential Loss of ANY two barriers.

Operating Mode Applicability:

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

Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.

Basis:

Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the Site Area Emergency classification level, each barrier is weighted equally.

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNFrlON CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:

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

Operating Mode Applicability:

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

Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.

Basis:

Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the Alert classification level, Fuel Cladding and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Cladding or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Cladding or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.

  • Basis Reference(s):
1. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY -

FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:

RCS Activity Operating Mode Applicability:

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

LOSS Coolant activity> 300 uCi/gm 1-131 dose equivalent.

Basis:

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

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a

  • sample-related threshold *is included as a backup to other indications .

There is no Potential Loss threshold associated with RCS Activity.

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • -* . RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Initiating Condition:

FC2 RPV Water Level Operating Mode Applicability:

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

LOSS

1. SAOG entry required.

POTENTIAL LOSS

2. RPV water level cannot be restored and maintained> 0 inches (TAF).

OR

3. RPV water level cannot be determined.

Basis:

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the EOP/SAOG program.

Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for SAOG entry. This is identified in the BWROG EPGs/SAGs when the phrase, "Enter SAOGs," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "SAOG entry required," also accommodates the EOP need to enter SAOGs when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.

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

The RPV water level threshold is the same as RCS Barrier RC2 Loss threshold. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

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

James*A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)

Basis (cont):

choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

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

  • considered reached as soon as it is apparent that the top of active fuel cannot be attained.

Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

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

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-2 RPV Control
3. EOP-7 RPV Flooding
4. EOP-3 Failure to Scram
5. NEI 99-01 Rev 6, Table 9-F-2

-.- James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear FC5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS Drywell radiation monitor reading> 1.8E+03 R/hr (1800 R/hr).

Basis:

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS

  • Barrier RCS Loss Threshold since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two thresholds appropriately escalates the emergency classification level to a Site Area Emergency.

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

Basis Reference(s):

1. EP-EAL-0715, Criteria for Choosing Containment Radiation values Indicating: loss of fuel clad and potential loss of containment for Fitzpatrick Nuclear Power Station
2. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS

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

POTENTIAL LOSS

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

Basis:

Loss Threshold #1 Basis This threshold addresses any other factors that are to be used by the Emergency

  • Director in determining whether the Fuel Clad Barrier is lost.

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

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex

- - - - -REC0GNITION-CATEGORY-F1SSION PRODUCT BARRIER DEGRADATION Exelon Nuclear RC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS

1. RPV water level cannot be restored and maintained > 0 inches (TAF)

OR

2. RPV water level cannot be determined.

Basis:

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the EOP/SAOG program.

This water level corresponds to the Top of Active Fuel (TAF) and is used in the EOPs to

  • indicate challenge to core cooling .

The RPV water level threshold is the same as Fuel Clad Barrier FC2 Potential Loss threshold. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

. -~* RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)

Basis (cont):

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

Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

  • There is no RCS Potential Loss threshold associated with RPV Water Level.

Basis Reference(s):

1.

2.

EP-1 EOP Entry and Use EOP-2 RPV Control

3. EOP-7 RPV Flooding
4. EOP-3 Failure to Scram
5. TSG-1 Parameter Assessment
6. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- - RECOGNl:rlON CATEGORY -- -- ----

FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:

Primary Containment Pressure/ Conditions Operating Mode Applicability:

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

LOSS

1. a. Primary Containment pressure> 2.7 psig.

AND

b. Primary Containment pressure rise is due to RCS leakage Basis:

The > 2.7 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

The second threshold condition focuses the fission product barrier loss threshold on a failure of the RCS instead of the non-LOCA malfunctions that may adversely affect

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

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-2 RPV Control
3. EOP-4 Primary Containment Control, Entry Conditions
4. FSAR Update Chapter 6 Emergency Core Cooling Systems
5. NEI 99-01 Rev 6, Table 9-F-2

l!

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- -- ~ -RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:

RCS Leak Rate Operating Mode Applicability:

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

LOSS

1. UNISOLABLE Main Steam Line (MSL), HPCI, RWCU, RCIC, or Feedwater line break.

OR

2. Emergency RPV Depressurization is required.

POTENTIAL LOSS

3. UNISOLABLE primary system leakage that results in EITHER of the following:
a. Secondary Containment area temperature > EOP-5 Maximum Normal Operating Limit.
  • Basis:

OR

b. Secondary Containment area radiation > EOP-5 Maximum Normal Operating Limit.

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Classification of a system break over system leakage is based on information available to the Control Room from the event. Indications that should be considered are:

  • Reports describing magnitude of steam or water release.
  • Use of system high flow alarms/ indications, if available,
  • Significant changes in makeup requirements,
  • Abnormal reactor water level changes in response to the event.

The use of the above indications provides the Control Room the bases to determine that the on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis (cont):

Loss Threshold #1 Basis Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, the RCS barrier Loss threshold is met.

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.

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

Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Radiation values in areas that contain piping associated with main steam line, RCIC, HPCI, Feedwater, RWCU, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is considered to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

--* RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis (cont):

Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold

  1. 1 (following automatic or manual isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Basis Reference(s):

1. FM-29A Main Steam System Flow Diagram
2. FM-29B Main Steam System Flow Diagram
3. FM-25A High Pressure Coolant Injection System Flow Diagram
4. FM-22A Reactor Core Isolation Cooling System Flow Diagram
5. FM-34A Feedwater System Flow Diagram
6. EP-1 EOP Entry and Use
7. EOP-2 RPV Control
8. EOP-3 Failure to Scram
9. EOP-4 Primary Containment Control
10. EOP-5 Secondary Containment Control
11. EOP-6 Radioactivity Release Control
12. EOP-7 RPV Flooding
13. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex

- - RECG>GNITION CATEGORY- -

FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear RCS Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS Drywell radiation monitor reading > 63 R/hr.

Basis:

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

There is no RCS Potential Loss threshold associated with Primary Containment Radiation .

  • Basis Reference(s):

1.

2.

EP-EAL-0515, Criteria for Choosing Drywell Radiation Monitor Reading Indicative of Loss of the RCS Barrier for Fitzpatrick Station NEI 99-01 Rev 6, Table 9-F-2

  • -~

James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear RC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS

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

POTENTIAL LOSS

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

Basis:

Loss Threshold #1 Basis This threshold addresses any other factors that are to be used by the Emergency

  • Director in determining whether the RCS Barrier is lost.

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

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

POTENTIAL LOSS SAOG entry required.

Basis:

The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV Water Level. The Potential Loss requirement for entry into the Severe Accident Procedures (SAOGs) indicates adequate core cooling cannot be restored and maintained and that core damage is possible. Entry into SAOGs is in response to the inability to restore and maintain adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential

  • for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-2 RPV Control
3. EOP-3 Failure to Scram
4. EOP-7 RPV Flooding
5. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Primary Containment Pressure/ Conditions Operating Mode Applicability:

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

LOSS

1. UNPLANNED rapid drop in Primary Containment pressure following Primary Containment pressure rise.

OR

2. Primary Containment pressure response not consistent with LOCA conditions.

POTENTIAL LOSS

3. Primary Containment pressure > 56 psig.

OR

4. a. Primary Containment hydrogen concentration ~ 6% .

AND

b. Primary Containment oxygen concentration ~ 5%.

OR

5. Heat Capacity Temperature Limit (HCTL) (EOP-11) exceeded.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Loss Threshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. A pressure suppression bypass path would not be an indication of a containment breach.

July 2019 JAF 2-66 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex

- - - RECOGNITION-CATEGORY -

FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear CT3 (cont)

Basis (cont):

Potential Loss Threshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.

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

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

Basis Reference(s):

1. FSAR Update Section 5.2.3
2. EOP-4 Primary Containment Control
3. UFSAR 14.6.1.3.3
4. BWROG EPG/SAG Revision 3, Sections PC/G
5. FSAR section 5.2.3.14
6. FSAR Table 7.3-6
7. BWROG EPG/SAG Revision 3, Section 18
8. EOP-11 EOP and SAOG Graphs
9. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- - -- RECOGNITION CATEGORY - - - -

FISSION PRODUCT BARRIER DEGRADATION CT5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

POTENTIAL LOSS Drywell radiation monitor reading > 1.8E+04 R/hr (18,000 R/hr).

Basis:

There is no Loss threshold associated with Primary Containment Radiation.

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

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20%

in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.

Basis Reference(s):

1. EP-EAL-0715, Criteria for Choosing Containment Radiation values Indicating: loss of fuel clad and potential loss of containment for Fitzpatrick Nuclear Power Station
2. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

__ RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:

Primary Containment Isolation Failure Operating Mode Applicability:

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

LOSS

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal.

OR

2. Intentional Primary Containment venting or purging per EOPs or SAOGs due to accident conditions.

OR

3. UNISOLABLE primary system leakage that results in EITHER of the following:
a. Secondary Containment area temperature > EOP-5 Maximum Safe Operating
  • Limit.

OR

b. Secondary Containment area radiation > EOP-5 Maximum Safe Operating Limit.

Basis:

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

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

Loss Threshold #1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).

Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE Main Steam Line, or RCIC, HPIC, Feedwater line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths.

Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.

July 2019 JAF 2-69 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system.

These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

Loss Threshold #2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed .

Loss Threshold #3 Basis The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is considered to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in

  • conjunction with other indications (e.g. room flooding, high area temperatures, reports of July 2019 JAF 2-70 EP-AA-1014 Addendum 3 (Rev. 1)

.-

  • James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear CT6 (cont)

Basis (cont):

steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Basis Reference(s):

1. EOP-4 Primary Containment Control
2. EP-6 Post accident Containment Venting and Gas Control
3. EP-1 EOP Entry and Use
4. EOP-5 Secondary Containment Control
5. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex

- -RECOGNITION-CA-"TEGORY-- - - -

FISSION PRODUCT BARRIER DEGRADATION Exelon Nuclear CT7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS

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

POTENTIAL LOSS

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

Basis:

Loss Threshold #1 Basis:

This threshold addresses any other factors that are to be used by the Emergency

  • Director in determining whether the Containment Barrier is lost.

Potential Loss Threshold #2 Basis:

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

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex RECOGNITl9N-CAT'EGORY SYSTEM MALFUNCTIONS Exelon Nuclear MG1 Initiating Condition:

Prolonged loss of all offsite and all onsite AC power to emergency buses.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of ALL offsite and onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. EITHER of the following:
a. Restoration of at least one 4160 V emergency bus 10500 or 10600 in

< 4 _hours is not likely .

  • Basis:

OR

b. RPV water level cannot be restored and maintained > -19 inches (MSCRWL).

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

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

James A. FitzPatrick Nuclear Power Plant Annex

- RECOGNITION CATEGORY- --

SYSTEM MALFUNCTIONS Exelon Nuclear 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.

If mitigating strategies are effective in reestablishing emergency power to any of the buses listed, within the specified time, then declaration of this EAL is not warranted.

This EAL is not concerned with the source of the power as much as the loss of power to the listed buses.

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. Misc. Calculation JAF-CALC-89-012 "Determination of Required SBO Coping Duration Per NUMARC 8700"
2. OP-44 115 KV System
3. Drawing 71-002 AC Distribution
4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer
7. JAFNPP Plant-Specific Technical Guideline (PSTG)
8. EOP-2 RPV Control
9. EOP-7 RPV Flooding
10. AOP-49 Station Blackout
11. NEI 99-01 Rev 6, SG1

James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear 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 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 onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 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.

If mitigating strategies are effective in reestablishing emergency power to any of the buses listed, within the specified time, then declaration of this EAL is not warranted.

This EAL is not concerned with the source of the power as much as the loss of power to the listed buses.

Escalation of the emergency classification level would be via ICs RG1, FG1, MG1, or MG2 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- RECOGNITION C-ATEGORY- - --

SYSTEM MALFUNCTIONS

  • MS1 (cont)

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-45 345 KV System
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer
7. AOP-49 Station Blackout
8. NEI 99-01 Rev 6, SS1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

--.- Initiating Condition:

RECOGNITION CAIEGORY-SYSTEM MALFUNCTIONS MA1 Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. AC power capability to 4160 V emergency buses 10500 and 10600 reduced to only one of the following power sources for~ 15 minutes.
  • Reserve Station Transformer T-2
  • Reserve Station Transformer T-3
  • EDGA
  • EOG B
  • EDGC
  • EOG D
  • Main Generator via T-4 AND
2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC MU1.

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.

July 2019 JAF 2-77 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex

-- -RECOGNITION CATEGORY-SYSTEM MALFUNCTIONS Exelon Nuclear MA1 (cont)

Basis (cont):

  • A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
  • Loss of offsite power and loss of all emergency power sources (e.g. onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

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. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-45 345 KV System
5. OP-22 Diesel Generator Emergency Power
  • 6.

7.

OP-45A Backfeeding Normal Station Service Transformer NEI 99-01 Rev 6, SA 1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

.-- RECOGNIIION CAIEGOR¥ SYSTEM MALFUNCTIONS MU1 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

Loss of ALL offsite AC power capability to 4160 V emergency buses 10500 and 10600 for~ 15 minutes.

  • Reserve Station Transformer T-2
  • Reserve Station Transformer T-3
  • Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses.

This condition represents a potential reduction in the level of safety of the plant.

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. (e.g. unit cross-tie breakers)

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC MA 1.

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-45 345 KV System
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer

-.~- ~

James A. FitzPatrick Nuclear Power Plant Annex

~RECOGNITION GAl"EGORY---

SYSTEM MALFUNCTIONS Exelon Nuclear MG2 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of ALL offsite and on site AC power to 4160 V emergency buses 10500 and 10600.

AND

2. Voltage is < 105 VDC on Vital DC buses 71 BCB-2A and 71 BCB-2B.

AND

3. ALL AC and Vital DC power sources in EALs #1 and #2 have been lost for

~ 15 minutes.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when all EAL conditions are met.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- RECO-GNITION CATE-GORY __ _

SYSTEM MALFUNCTIONS MG2 (cont)

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-22 Diesel Generator Emergency Power
5. OP-45A Backfeeding Normal Station Service Transformer
6. JAFNPP Plant-Specific Technical Guideline (PSTG)
7. EOP-2 RPV Control
8. EOP-7 RPV Flooding
9. AOP-49 Station Blackout
10. Drawing S71-068
11. OP-43A 125 voe System
12. ARP 09-8-1-20 125 voe Batt A Volt Lo ARP 09-8-1-23 125 voe Batt B Volt Lo 13 .
14. AOP-45 Loss of DC Power System 'A'
15. AOP-46 Loss of DC Power System 'B'
16. NEI 99-01 Rev 6, SG8

James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION-CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MS2 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

Voltage is < 105 VDC on Vital DC buses 71 BCB-2A and 71 BCB-2B 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. Drawing S71-068
2. OP-43A 125 voe System
3. ARP 09-8-1-20 125 voe Batt A Volt Lo
4. ARP 09-8-1-23 125 voe Batt B Volt Lo
5. AOP-45 Loss of DC Power System 'A'
6. AOP-46 Loss of DC Power System 'B'
7. NEI 99-01 Rev 6, SS8

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 Initiating Condition:

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

Operating Mode Applicability:

1, 2 Emergency Action Level (EAL):

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

~2.5%.

AND

2. ALL manual I ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power~ 2.5%.

AND

3. EITHER of the following conditions exist:
  • RPV water level cannot be restored and maintained > -19 inches (MSCRWL).

OR Basis:

Heat Capacity Temperature Limit (HCTL) (EOP-11) exceeded.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection, are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

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

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

Escalation of the emergency .classification level would be via IC RG1 or FG1.

July 2019 JAF 2-83 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Basis Reference(s):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS Exelon Nuclear MS3 (cont)

1. FSAR Update Section 7.2
2. EOP-3 Failure to Scram
3. EOP-2 RPV Control
4. EOP-4 Primary Containment Control
5. EOP-7 RPV Flooding
6. EOP-11 EOP and SAOG Graphs
7. NEI 99-01 Rev 6, SS5

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the Reactor Control Console 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. This action does not include manually driving in control rods or implementation of boron injection strategies.
1. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power.=:. 2.5%.

AND

2. Manual / ARI actions taken at the Reactor Control Console are not successful in shutting down the reactor as indicated by Reactor Power.=:. 2.5%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the Reactor Control Console to rapidly 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 scram, ARI). 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".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • Basis (cont):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 (cont)

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

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

Basis Reference(s):

1. EP-3 Backup Control Rod Insertion
2. EOP-3 Failure to Scram
3. EOP-2 RPV Control

James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION -CATEGOR¥-

SYSTEM MALFUNCTIONS Exelon Nuclear MU3 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor.

Operating Mode Applicability:

1, 2 Emergency Action Level (EAL):

Note:

  • A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core. This action does not include manually driving in control rods or implementation of boron injection strategies.
1. a. Automatic scram did not shutdown the reactor as indicated by Reactor Power~ 2.5%.

AND

b. Subsequent manual / ARI action taken at the Reactor Control Console is successful in shutting down the reactor as indicated by Reactor Power < 2.5% .

OR

2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power~ 2.5%.

AND

b. EITHER of the following:
1. Subsequent manual / ARI action taken at the Reactor Control Console is successful in shutting down the reactor as indicated by Reactor Power < 2.5%.

OR

2. Subsequent automatic scram/ ARI is successful in shutting down the reactor as indicated by Reactor Power< 2.5%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the Reactor Control Console or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)

Basis (cont):

EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the Reactor Control Console to shutdown the reactor (e.g., initiate a manual reactor scram/ARI). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

EAL#2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take other manual actions on the Reactor Control Console to shutdown the reactor (e.g., initiate a manual reactor scram / or initiating ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram, including ARI, is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the Reactor Control 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 scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the Reactor Control Console".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, .etc. If subsequent operator manual actions taken at the Reactor Control 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 FA 1. Absent the plant conditions needed to meet either IC MA3 or FA 1, an Unusual Event declaration is appropriate for this event.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

  • If the signal generated as a result of plant work causes a plant transient that creates a real condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • Basis (cont):

RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)

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

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-3 Failure to Scram
3. EOP-2 RPV Control
4. AOP-1 Reactor Scram
5. Technical Specifications section 3.3.1.1 RPS Instrumentation
6. NEI 99-01 Rev 6, SUS

James A. FitzPatrick Nuciear Power Plant Annex Exelon Nuclear RECOGNIIION CATEGORY- -

SYSTEM MALFUNCTIONS MA4 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for~ 15 minutes.

Table M1 Control Room Parameters

  • Reactor Power
  • RPV Water Level
  • Torus Level
  • Torus Temperature AND
2. ANY Table M2 transient in progress.

Table M2 Significant Transients

  • Auto/Manual runback > 25% thermal reactor power
  • Electric load rejection > 25% full electric load
  • Thermal Power oscillations > 10% (peak to peak)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- -- -- -RECO-GNITION CATE-GORY --

SYSTEM MALFUNCTIONS MA4 (cont)

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier* challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for any of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, computer point, digital, recorder source, or equivalent (e.g. camera) within the Control Room.

  • An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RS1 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 (cont)

Basis Reference(s):

1. FSAR Update Section 7.16
2. FSAR Update Section 7.19
3. EOP-2 RPV Control
4. EOP-3 Failure to Scram
5. EOP-4 Primary Containment Control
6. EOP-5 Secondary Containment Control
7. EOP-6 Radioactivity Release Control
8. EOP-7 RPV Flooding
9. NEI 99-01 Rev 6, SA2

-~. James A. FitzPatrick Nuclear Power Plant Annex RECOGNIIION-CAIEGOR't'---

SYSTEM MALFUNCTIONS Exelon Nuclear MU4 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for~ 15 minutes.

Table M1 Control Room Parameters

  • Reactor Power
  • RPV Water Level
  • Basis:

Primary Containment Pressure Torus Level Torus Temperature UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for any of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital, recorder source, or equivalent (e.g. camera) within the Control Room .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4 (cont)

Basis (cont):

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

  • Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC MA4.

Basis Reference(s):

1. FSAR Update Section 7.16
2. FSAR Update Section 7.19
3. NEI 99-01 Rev 6, SU2

.-- James A. FitzPatrick Nuclear Power Plant Annex Initiating Condition:

RECOGNIIION CATEGORY-SYSTEM MALFUNCTIONS Exelon Nuclear MAS Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.

Operating Mode Applicability:

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

Note:

  • This EAL is only applicable to SAFETY SYSTEMs having two (2) or more trains.
  • If the affected SAFETY SYSTEM train was already inoperable before the -

hazardous event occurred, then this emergency classification is not warranted.

  • If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
  • If a hazardous event occurs and it is determined that the conditions of MA5 are not met, then assess the event via HU3, HU4, or HU6.
1. a. The occurrence of ANY of the following hazardous events:

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

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

AND

c. EITHER of the following:
  • Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

OR

  • Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode .

July 2019 JAF 2-95 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAS (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.

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train 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 SAFETY SYSTEM train .

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS required for the current operating mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. In order to provide the appropriate context for consideration of an Alert classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has degraded performance for criteria 1.b of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance address damage to a SAFETY SYSTEM train that 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 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAS (cont)

Basis (cont):

Operators will make a determination of VISIBLE DAMAGE 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or RS1.

If a hazardous event occurs and 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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

RECOGNIIION_CAIEGORY SYSTEM MALFUNCTIONS MU6 Initiating Condition:

RCS leakage for 15 minutes or longer.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for~ 15 minµtes.

OR

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

OR

3. Leakage from the RCS to a location outside the Drywell > 25 gpm for~ 15 minutes.
  • Basis:

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

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. The assessment of this EAL may be based on the results of RCS leak rate calculation that may be necessary to ascertain whether the EAL has been exceeded. In this case, the 15-minute declaration period starts with the availability of the RCS leak rate calculation results that show the EAL to be exceeded (i.e., this is the time that the EAL information is first available).

These three EALs thus apply to leakage into the containment, a secondary-side system

  • or a location outside of containment.

July 2019 JAF 2-98 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY-SYSTEM MALFUNCTIONS MU6 (cont)

Basis (cont):

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

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

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category R or F.

Basis Reference(s):

1. FSAR Update Section 4.10
2. NEI 99-01 Rev 6, SU4

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- - RECOGNITION-CATEGORY- ~ -

SYSTEM MALFUNCTIONS MU7 Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

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

1. Loss of ALL Table M3 onsite communication capabilities affecting the ability to perform routine operations.

OR

2. Loss of ALL Table M3 offsite communication capabilities affecting the ability to perform offsite notifications.

OR

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

Table M3 Communication Capabilities

  • System Page/Party System (Gaitronics)

Control Room/Portable Radio Plant Telephones (all VOiP, switched, non-switched)

Onsite X

X X

Offsite X

NRC X

Installed Out-of-Plant Cellular Phones X X X Plant Satellite Phones (Installed in CR and deployable) X X RECS X Dedicated Phone Lines (ENS) X X HPN and FTS 2001 X X Basis:

This IC addresses a significant loss of onsite, offsite, or NRC communication 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.) .

  • July 2019 JAF 2-100 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- -- RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)

Basis (cont):

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

EAL #2 Basis Addresses a total loss of the communication methods used to notify all Offsite Response Organizations (OROs) of an emergency declaration. The Offsite Response Organizations (OROs) referred to here are listed in procedure EP-CE-114-100-F-05, JAF Notification Fact Sheet .

EAL #3 Basis Addresses a total loss of the communication methods used to notify the NRG of an emergency declaration.

Basis Reference(s):

1. NY State Emergency Operations Center
2. NY State Warning Point
3. Alternate State Warning Point
4. State Department of Health
5. SEMO Regional Office
6. Oswego County EOG
7. Oswego County E-911 Center (Warning Point)
8. Nine Mile Point Control Rooms
9. Nine Mile Point TSC and EOF
10. JAFNPP Control Room
11. JAFNPP TSC
12. JAFNPP EOF
13. SEMO Technical Resources 14 . NEI 99-01 Rev 6, SU6
  • July 2019 JAF 2-101 EP-AA-1014 Addendum 3 (Rev. 1)

-.-- James A. FitzPatrick Nuclear Power Plant Annex

- - ---- --- -- -- -- RECOGNITIONGA1"EG0RY --- -~- -

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Initiating Condition:

Exelon Nuclear CA1 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

Operating Mode Applicability:

4,5, D Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of ALL offsite and onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 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.

If mitigating strategies are effective in reestablishing emergency power to any of the buses listed, within the specified time, then declaration of this EAL is not warranted.

This EAL is not concerned with the source of the power as much as the loss of power to the listed buses .

  • July 2019 JAF 2-102 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

. -- - **- . -RECOGNITION-CATEGORY**

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)

Basis (cont):

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.

Basis Reference(s):

1. Drawing 71-002 AC Distribution
2. OP-44 115 KV System
3. OP-45 345 KV System
4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer
7. NEI 99-01 Rev 6, CA2
  • July 2019 JAF 2-103 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

---.---- -~ _ _____ RECOGNITION CATEGORY-COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU1 Initiating Condition:

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

Operating Mode Applicability:

4,5, D Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. AC power capability to 4160 V emergency buses 10500 and 10600 reduced to only one of the following power sources for~ 15 minutes.
  • Reserve Station Transformer T-2
  • Reserve Station Transformer T-3
  • EDGA
  • EOG B
  • EDGC
  • EOG D AND
2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

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

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

July 2019 JAF 2-104 EP-AA-1014 Addendum 3 (Rev. 1)

.-- James A. FitzPatrick Nuclear Power Plant Annex Basis (cont):

-- --- ---- ~ RECOGNITION CAT-EGORY -- ----- ---

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Exelon Nuclear

~- -- - --

CU1 (cont)

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

  • A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
  • Loss of offsite power and loss of all emergency power sources (e.g. onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

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 CA 1 .

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
  • 3.

4.

5.

6.

OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution OP-45 345 KV System OP-22 Diesel Generator Emergency Power OP-45A Backfeeding Normal Station Service Transformer

7. NEI 99-01 Rev 6 CU2
  • July 2019 JAF 2-105 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- - - - --- - - - - - - - ---------RECOGNll'ION-CAl'EGOR't---

  • COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:

Hazardous event affecting SAFETY SYSTEM required for the current operating mode.

Operating Mode Applicability:

4, 5 Emergency Action Level (EAL):

Note:

  • This EAL is only applicable to SAFETY SYSTEMs having two (2) or more trains.
  • If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred, then this emergency classification is not warranted.
  • If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
  • If a hazardous event occurs and it is determined that the conditions of CA2 are not met, then assess the event via HU3, HU4, or HU6.
1. a. The occurrence of ANY of the following hazardous events:

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

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

AND

c. EITHER of the following:
  • Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

OR

  • Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode .
  • July 2019 JAF 2-106 EP-AA-1014 A:ddendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- - ----- -- -- -RECOGNFFION-GA~EGQRY ~ -- - -- -- --- ---

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (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.

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train 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 SAFETY SYSTEM train .

  • This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS required for the current operating mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. In order to provide the appropriate context for consideration of an Alert classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has degraded performance for criteria 1.b of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance address damage to a SAFETY SYSTEM train that 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 .

  • July 2019 JAF 2-107 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- ---~----

RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (cont)

Basis (cont):

Operators will make a determination of VISIBLE DAMAGE 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or RS1.

If a hazardous event occurs and the EAL conditions of MAS are not met then assess the event via HU3, HU4, or HU6.

Basis Reference(s):

1. NEI 99-01, Rev 6 CA6
  • July 2019 JAF 2-108 EP-AA-1014 Addendum 3 (Rev. 1)

~.--- James A. FitzPatrick Nuclear Power Plant Annex Initiating Condition:

RECOGNl'T'ION-CA'T'EGOR'i'------

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Exelon Nuclear CU3 Loss of Vital DC power for 15 minutes or longer.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

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

Voltage is < 105 VDC on required Vital DC buses 71 BCB-2A and 71 BCB-28 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 Bis 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.

  • July 2019 JAF 2-109 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 (cont)

Basis Reference(s):

1. Drawing S71-068
2. OP-43A 125 voe System
3. ARP 09-8-1-20 125 voe Batt A Volt Lo
4. ARP 09-8-1-23 125 voe Batt AB Volt Lo
5. AOP-45 Loss of DC Power System 'A'
6. AOP-46 Loss of DC Power System 'B'
7. NEI 99-01 Rev 6, CU4
  • July 2019 JAF 2-110 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- --- RECOGNITION-CA"FEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

4,5, D Emergency Action Level (EAL):

1. Loss of ALL Table C1 onsite communication capabilities affecting the ability to perform routine operations.

OR

2. Loss of ALL Table C1 offsite communication capabilities affecting the ability to perform offsite notifications.

OR

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

Table C1 Communication Capabilities

  • System Page/Party System (Gaitronics)

Control Room/Portable Radio Plant Telephones (all VOiP, switched, non-switched)

Onsite X

X X

Offsite X

NRC X

Installed Out-of-Plant Cellular Phones X X X Plant Satellite Phones (Installed in CR and deployable) X X RECS X Dedicated Phone Lines (ENS) X X HPN and FTS 2001 X X Basis:

This IC addresses a significant loss of onsite, offsite, or NRC communication 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.) .

July 2019 JAF2-111 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nu'clear

- -- - - - - --- - -~ --~ -

RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 (cont)

Basis (cont):

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

EAL #2 Basis Addresses a total loss of the communication methods used to notify all Offsite Response Organizations (OROs) of an emergency declaration. The Offsite Response Organizations (OROs) referred to here are listed in procedure EP-CE-114-100-F-05, JAF Notification Fact Sheet.

EAL #3 Basis Addresses a total loss of the communication methods used to notify the NRC of an emergency declaration.

  • Basis Reference(s):
1. NY State Emergency Operations Center
2. NY State Warning Point
3. Alternate State Warning Point
4. State Department of Health
5. SEMO Regional Office
6. Oswego County EOC
7. Oswego County E-911 Center (Warning Point)
8. Nine Mile Point Control Rooms
9. Nine Mile Point TSC and EOF
10. JAFNPP Control Room
11. JAFNPP TSC
12. JAFNPP EOF
13. SEMO Technical Resources 14 . NEI 99-01 Rev 6, CU5
  • July 2019 JAF 2-112 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

---.- RECOGNfflON CATEGOR¥ COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS

-~- -

CA5 Initiating Condition:

Inability to maintain the plant in cold shutdown.

Operating Mode Applicability:

4, 5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.
1. UNPLANNED rise in RCS temperature > 212 °F for> Table C2 duration.

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

Established 20 minutes*

Not Intact Not Established 0 minutes

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

OR

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

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions .

  • July 2019 JAF 2-113 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)

Basis (cont):

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise.

The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the

  • temperature rise without a substantial degradation in plan~ safety .

Finally, in the case where there is a rise in RCS temperature, the RCS is not intact , and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL #2 provides a pressure-based indication of RCS heat-up.

Escalation of the emergency classification level would be via IC CS6 or RS1.

Basis Reference(s}:

1. Technical Specifications Table 1.1-1
2. AOP-30 Loss of Shutdown Cooling
3. OP-13D RHR-Shutdown Cooling
4. Technical Specifications Section 3.61.1 and 3.6.4.1
5. NEI 99-01 Rev 6, CA3
  • July 2019 JAF 2-114 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

_ ___ - -- RECOGNll'ION CATEGORY -

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 Initiating Condition:

UNPLANNED rise in RCS temperature Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.
1. UNPLANNED rise in RCS temperature> 212 °F.

OR

2. Loss of the following for~ 15 minutes .
  • Basis:

ALL RCS temperature indications AND ALL RPV water level indications UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature

  • limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA5.

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.) .

James A. FitzPatrick Nuclear Power Plant Annex

_ __ RECOGNITION-CAl'EGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Exelon Nuclear CU5 (cont)

Basis (cont):

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

EAL#1 Basis This 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 Basis This reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. .

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA6 based on an inventory loss or IC CA5 based on exceeding plant configuration-specific time criteria.

Basis Reference(s):

1. Technical Specifications Table 1.1-1
2. AOP-30 Loss of Shutdown Cooling
3. Drawing S02-069
4. NEI 99-01 Rev 6, CU3
  • July 2019 JAF 2-116 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- - -- -- -- -- - -- -RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Initiating Condition:

Loss of RPV inventory affecting fuel clad integrity with c~ntainment challenged.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. a. RPV water level < 0 inches (T AF) for~ 30 minutes.

AND

b. ANY Table C4 Containment Challenge Indication.

OR

2. a. RPV water level cannot be monitored for~ 30 minutes.

AND

b. Core uncovery is indicated by ANY of the following:
  • Table C3 indication of a sufficient magnitude to indicate core uncovery.

OR

~ 3 R/hr.

AND

c. ANY Table C4 Containment Challenge Indication .
  • July 2019 JAF 2-117 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

.--- - -~ - - RECOGNITION CA"J'.EGORY -- -- --

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Emergency Action Level (EAL) (cont):

Table C3 Indications of RCS Leakage

  • UNPLANNED Drywell equipment drain sump level rise*
  • UNPLANNED Drywell floor drain sump level rise*
  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory.

Table C4 Containment Challenge Indications Primary Containment Hydrogen Concentration UNPLANNED rise in primary containment pressure CONTAINMENT CLOSURE not established*

~ 6% and Oxygen ~

Secondary Containment area radiation > ANY Maximum Safe Operating Limit 5%

(EOP-5)

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

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

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

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions .

  • July 2019 JAF 2-118 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- -- -~ - --RECOGNITION-CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Basis (cont):

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guidelines (PAG) exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment

-atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity .

  • In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access.

During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

EAL#2 Basis The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

  • July 2019 JAF 2-119 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- --- -- RECOGNITION-CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Basis (cont):

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.

Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G
2. EOP-4a Primary Containment Gas Control
3. FSAR section 5.2.3.14
4. FSAR Update Table 5.2-1
5. Technical Support Guideline-1 (TSG-1) Parameter Assessment
6. FSAR Update Section 4.10.3
7. OP-13D RHR-Shutdown Cooling
8. EOP-5 Secondary Containment Control
9. Technical Specifications Sections 3.6.1.1 and 3.6.4.1
10. EOP-2 RPV Control
11. EP-EAL-0506 Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Fitzpatrick Station 12 . NEI 99-01 Rev 6, CG1
  • July 2019 JAF 2-120 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ~---------------

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 Initiating Condition:

Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. With CONTAINMENT CLOSURE not established, RPV water level

< 120.5 inches.

OR

2. With CONTAINMENT CLOSURE established, RPV water level< 0 inches (TAF).

OR

3. a. RPV water level cannot be monitored for~ 30 minutes AND
b. Core uncovery is indicated by ANY of the following:
  • Table C3 indication of a sufficient magnitude to indicate core uncovery.

OR

  • July 2019 JAF 2-121 EP-AA-1014 Addendum 3 (Rev. 1)

--. James A. FitzPatrick Nuclear Power Plant Annex RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Emergency Action Level (EAL) (cont):

Exelon Nuclear CS6 (cont)

Table C3 Indications of RCS Leakage

  • UNPLANNED Drywell equipment drain sump level rise*
  • UNPLANNED Drywell floor drain sump level rise*
  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

EAL #1 and #2 Basis The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs #1 and #2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

  • July 2019 JAF 2-122 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- - - --- -~ -- --- --

~- ~ -RECOGN-ITION CATEGORY ~ ----

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)

Basis (cont):

EAL #3 Basis The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It. also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess

  • Shutdown Management.

Escalation of the emergency classification level would be via IC CG6 or RG1 .

  • July 2019 JAF 2-123 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 {cont)

Basis Reference(s):

1. Technical Support Guideline-1 (TSG-1) Parameter Assessment
2. FSAR Update Section 4.10.3
3. OP-13D RHR-Shutdown Cooling
4. EOP-5 Secondary Containment Control
5. Technical Specifications Sections 3.6.1.1 and 3.6.4.1
6. EOP-2 RPV Control
7. EP-EAL-0506 Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Fitzpatrick Station
8. NEI 99-01 Rev 6, CS1
  • I
  • July 2019 JAF 2-124 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- - -- ---- --- -- ----~

-- --- ~ -

.~.. - - - -RECOGNlTICfN CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:

Loss of RPV inventory.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of RPV inventory as indicated by level < 126.5 inches.

OR

2. a. RPV water level cannot be monitored for~ 15 minutes.

AND

b. Loss of RPV inventory per Table C3 indications .

Table C3 Indications of RCS Leakage UNPLANNED Drywell equipment drain sump level rise*

UNPLANNED Drywell floor drain sump level rise*

  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses 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 .

  • July 2019 JAF 2-125 EP-AA-1014 Addendum 3 (Rev. 1)

- .- James A. FitzPatrick Nuclear Power Plant Annex Basis (cont):

RECOGNITION CATEGORY-COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS Exelon Nuclear CA6 {cont)

EAL #1 Basis A lowering of water level below 126.5 inches indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery.

Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA5.

EAL #2 Basis The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS6 If the RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS6.

Basis Reference(s):

1. Technical Specifications Table 3.3.5.1.-1
2. Drawing S02-069
3. FSAR Update Section 4.10.3
4. OP-130 RHR-Shutdown Cooling
5. NEI 99-01 Rev 6, CA1
  • July 2019 JAF 2-126 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 Initiating Condition:

UNPLANNED loss of RPV inventory for 15 minutes or longer.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. UNPLANNED loss of reactor coolant results in the inability to restore and maintain RPV water level above the procedurally established lower limit for> 15 minutes.

OR

2. a. RPV water level cannot be monitored.

AND

b. Loss of RPV inventory per Table C3 indications.

Table C3 Indications of RCS Leakage

  • UNPLANNED Drywell equipment drain sump level rise*
  • UNPLANNED Drywell floor drain sump level rise*
  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory .
  • July 2019 JAF 2-127 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

____ _ - -- - RECOGNITION-CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 {cont)

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV water level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

The procedurally established lower limit is not an operational band established above the procedural limit to allow for operator action prior to exceeding the procedural limit, but it is the procedurally established lower limit.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that re.suits in water level lowering 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 This recognizes that the minimum required RPV water level can change several times

  • during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL #2 Basis This addresses a condition where all means to determine RPV water level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA6 or CA5 .

  • July 2019 JAF 2-128 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

-- - RECOGNITIONCATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)

Basis Reference(s):

1. Drawing S02-069
2. FSAR Update Section 4.10.3
3. OP-13D RHR-Shutdown Cooling
4. Technical Support Guideline-1 (TSG-1) Parameter Assessment
5. OP-658 Shutdown Operation
6. EOP-2 RPV Control
7. NEI 99-01, Rev. 6 CU1
  • July 2019 JAF 2-129 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

- ----- -- -- - -- -RECO-GNITIONCATEGORY -

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:

HOSTILE ACTION within the PROTECTED AREA.

Operating Mode Applicability:

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

A notification from the Security Supervisor that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area) .

  • HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program] .

July 2019 JAF 2-130 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 (cont)

Basis (cont):

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (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 and HG?.

Basis Reference(s):

1. JAFNPP Safeguards Contingency Plan
  • 2.

3.

AOP-70 Security Threat NEI 99-01 Rev 6, HS1

  • July 2019 JAF 2-131 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 Initiating Condition:

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

Operating Mode Applicability:

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

1. A validated notification from NRC of an aircraft attack threat< 30 minutes from the site.

OR

2. Notification by the Security Supervisor 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

  • July 2019 JAF 2-132 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont}

Basis (cont):

PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE.

Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72 .

EAL #1 Basis Addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with AOP-70, Security Threat.

EAL #2 Basis Applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Escalation of the emergency classification level would be via IC HS1 .

July 2019 JAF 2-133 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)

Basis Reference(s):

1. JAFNPP Safeguards Contingency Plan
2. AOP-70 Security Threat
3. NEI 99-01 Rev 6, HA1
  • July 2019 JAF 2-134 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:

Confirmed SECURITY CONDITION or threat.

Operating Mode Applicability:

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

1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.

OR

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

OR

3. Notification by the Security Supervisor 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 threaUcompromise to site security, threaUrisk to site

  • personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety.

Security events which do not meet one of these EALs are adequately addressed by the

  • July 2019 JAF 2-135 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 (cont)

Basis (cont):

requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, and HS1.

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 Offsite Response Organizations (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 AOP-70, Security 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 HA 1.

Basis Reference(s):

1. JAFNPP Safeguards Contingency Plan
2. AOP-70 Security Threat
3. SY-AA-101-132, Security Assessment and Response to Unusual Activities
4. NEI 99-01 Rev 6, HU1
  • July 2019 JAF 2-136 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition:

Inability to control a key safety function from outside the Control Room.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the Control Room.

AND

2. Control of ANY Table H1 key safety function is not reestablished in< 30 minutes.
  • Table H1 Safety Functions
  • Reactivity Control (ability to shut down the reactor and keep it shutdown)
  • RPV Water Level (ability to cool the core)
  • RCS Heat Removal (ability to maintain heat sink)

Basis:

The time period to establish control of the plant starts when either:

a. Control of needed safety functions 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 barrier within a relatively short period of time .

  • July 2019 JAF 2-137 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)

Basis (cont):

The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 30 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Escalation of the emergency classification level would be via IC FG1 or CG6.

Basis Reference(s):

1. AOP-43 Plant Shutdown from Outside the Control Room
2. NEI 99-01, Rev 6 HS6
  • July 2019 JAF 2-138 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2 Initiating Condition:

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

Operating Mode Applicability:

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

A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the Control Room.

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 t~e 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. AOP-43 Plant Shutdown from Outside the Control Room
2. NEI 99-01, Rev 6 HA6
  • July 2019 JAF 2-139 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition:

FIRE potentially degrading the level of safety of the plant.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • Escalation of the emergency classification level would be via IC CA2 or MAS.
1. A FIRE in ANY Table H2 area is not extinguished in< 15 minutes of ANY of the following FIRE detection indications:
  • Report from the field (i.e., visual observation)
  • Receipt of multiple (more than 1) fire alarms or indications
  • Field verification of a single fire alarm Table H2 Areas
  • Reactor Building (when inerted the Drywell is exempt)
  • Control Room I Relay Room I Cable Run Rooms / Cable Spreading Room
  • Electric Bays
  • Control Room AC Equipment Room
  • Control Room Chiller Room
  • Battery Rooms / Battery Room Corridor
  • Cable Tunnels

OR

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

AND

b. The existence of a FIRE is not verified in< 30 minutes of alarm receipt.

OR

  • July 2019 JAF 2-140 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Emergency Action Level {EAL) {cont):

3. A FIRE within the plant PROTECTED AREA not extinguished in < 60 minutes of the initial report, alarm or indication.

OR

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

Basis:

FIRE: Combustion characterized by heat and light. Sources of smoke such- as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) : A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

  • This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

EAL #1 Basis

  • The intent of the 15 minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarms, indication or report.

EAL #2 Basis This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30 minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30 minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed .

  • July 2019 JAF 2-141 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Basis (cont):

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30 minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30 minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

EAL #3 Basis In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60 minutes may also potentially degrade the level of plant safety.

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 July 2019 JAF 2-142 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Basis (cont):

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. FSAR Update Section 12.3
2. JAFNPP Safe Shutdown Analysis
3. NEI 99-01, Rev 6 HU4
  • July 2019 JAF 2-143 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition:

Seismic event greater than QBE levels.

Operating Mode Applicability:

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

Note:

  • For emergency classification if EAL # 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Emergency Director in~ 15 minutes of the event.
  • Escalation of the emergency classification level would be via IC CA2 or MAS
1. Seismic event > Operating Basis Earthquake (QBE) as determined by seismic monitoring system in accordance with AOP-14 Earthquake.

OR

2. When Seismic Monitoring Equipment is not available:
a. Control Room personnel feel an actual or potential seismic event.

AND

b. ANY one of the following confirmed in~ 15 minutes of the event:
  • The earthquake resulted in Modified Mercalli Intensity (MMI) ~ VI and occurred

~ 3.5 miles of the plant.

  • The earthquake was magnitude~ 5.0 and occurred i 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 a manner deemed appropriate by the Shift Manager or Emergency Director.
  • July 2019 JAF 2-144 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)

Basis:

EAL #1 Basis This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (QBE) 1 . An earthquake greater than an QBE but less than a Safe Shutdown Earthquake (SSE)2 should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an QBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

EAL #2 Basis 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 guidance recognizes that it may cause the site to declare an Unusual Event while another site, similarly affected but with readily available QBE 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.

1 An QBE 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.

July 2019 JAF 2-145 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)

Basis Reference(s):

1. FSAR Update Section 2.6 Engineering Seismology
2. AOP-14 Earthquake
3. US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions
4. NEI 99-01, Rev 6 HU2
  • July 2019 JAF 2-146 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAS 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 a Table H3 area.

Areas with Ent Area Reactor Building East Crescent Reactor Building West Crescent Reactor Building 272' Elevation Reactor Building 300' Elevation Relay Room Mode 3, 4, and 5

  • North Cable Room 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.

  • July 2019 JAF 2-147 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAS (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 preclude the ability to place shutdown cooling in service. The emergency classification is not contingent upon whether entry is actually necessary at the time. of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

An emergency declaration is not warranted if any of the following conditions apply.

  • The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
  • . The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
  • The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections) .

July 2019 JAF 2-148 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAS (cont)

Basis (cont):

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that generate smoke, that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

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

Basis Reference(s):

1. JAFNPP Safe Shutdown Analysis
2. NEI 99-01, Rev 6 HA5
  • July 2019 JAF 2-149 EP-AA-1014 Addendum 3 (Rev. 1)

James-A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HUS Initiating Condition:

Hazardous Event Operating Mode Applicability:

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

Note:

  • EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
  • Escalation of the emergency classification level would be via IC CA2 or MAS.
1. Tornado strike within thePROTECTED AREA.

OR

2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode .
  • OR 3, Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Intake Water Level > 255 feet.

OR

6. ESW intake bay water level~ 237 feet.

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.

July 2019 JAF 2-150 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)

Basis {cont):

EAL #1 Basis Addresse*s a tornado striking (touching down) within the Protected Area.

EAL #2 Basis Addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

EAL #3 Basis Addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.

EAL #4 Basis Addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy

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 The high lake level is based upon the revised design flood level for the screenwell interior walls and gates.

EAL #6 Basis The low level of~ 237 feet is selected since IAW AOP-56, Intake Water Level Trouble at 237.5 feet there is adequate time to reduce power (scraming the reactor), remove the second CW pump and trend intake water level as mitigating actions. The ESW pumps are declared inoperable (Alert threshold) at 236.5 feet, so 237 feet allows for mitigating action to be taken prior to declaration and is above the Alert threshold allowing for escalation between the Unusual Event and the Alert thresholds.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, Hor C .

  • July 2019 JAF 2-151 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)

Basis Reference(s):

1. FSAR Section 2.4.3
2. Safety Evaluation JAF-SE-93-034 "Evaluation of Maximum and Minimum Water Levels at Screenwell for Safe Operation of Class I Equipment"
3. NEI 99-01, Rev 6 HU3
4. AOP-56, Intake Water Level Trouble
  • July 2019 JAF 2-152 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG7 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY.

Operating Mode Applicability:

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

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

Basis:

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

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.

Basis Reference(s):

1. NEI 99-01, Rev 6 HG?
  • July 2019 JAF 2-153 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Initiating Condition:

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

Operating Mode Applicability:

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

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to

  • achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.

Basis Reference(s):

1. NEI 99-01, Rev 6 HS?
  • July 2019 JAF 2-154 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT.

Operating Mode Applicability:

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

Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts

  • that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety. -

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert.

Basis Reference(s):

1. NEI 99-01, Rev 6 HA?
  • July 2019 JAF 2-155 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU7 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.

Operating Mode Applicability:

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

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENT.

  • Basis Reference(s):
1. NEI 99-01, Rev 6 HU?
  • July 2019 JAF 2-156 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability:

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

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading> ANY Table E-1 values:

Table E-1 Radiation Reading Overpack Serial Overpack Average Overpack Overpack Number Surface Dose Serial Average Surface Rates mrem/hr Number Dose Rates (gamma+neutron) Hl:.STORM mrem/hr (gamma+neutron) 100S (XXX)

  • HI-STORM 100S S/N -15, 16, 17 80 on the side 20 on the top 32 at the inlet and outlet vent ducts S/N - 0186, 0187, 0188 S/N -0307, 0308, 0309, 220 on the side 40 on the top 0310, 0311, 0312, 0679, HI-STORM 100 on the side 600 on the side 0680, 0681, 1oos (232) 0682, 0683, 60 on the top 20 on the top S/N - 0169, 0690, 0691, 90 at the inlet and 0692, 0693, 0170,0171 outlet vent 0694, 0695 ducts

.Basis:

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

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) : A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage .

  • July 2019 JAF 2-157 EP-AA-1014 Addendum 3 (Rev. 1)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-H U1 (cont)

Basis (cont):

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The word cask, as used in this EAL, refers to the storage container in use at the site for dry storage of irradiated fuel. The issues of concern are the creation of a potential or actual release path to the environment, degradation of any fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSls are covered under ICs HU1 and HA1 .

  • Basis Reference(s):

1.

2.

NEI 99-01, Rev 6 E-HU1 ISFSI Certificate Of Compliance Amendment No.'s 0, 1, 2, 5, and 8 .

  • July 2019 JAF 2-158 EP-AA-1014 Addendum 3 (Rev. 1)

ATTACHMENT 3 Radiological Emergency Plan Document Revision EP-AA-1014 Addendum 3, Revision 2, "James A. FitzPatrick Nuclear Power Plant Emergency Action Levels"

EP-AA-1014 Addendum 3

~* Exelon Generation~ Revision 2 JAMES A. FITZPATRICK I. NUCLEAR POWER PLANT EMERGENCY ACTION LEVELS

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • REVISION HISTORY Rev. 1 Rev. 2 July 2019 July 2019

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • Section 1: Classification of Emergencies 1.1 General Section D of the Exelon Nuclear Standardized Emergency Plan divides the types of emergencies into four Emergency Classification Levels (ECLs). The four are the Unusual Event (UE), Alert, Site Area Emergency (SAE), and General Emergency (GE). These ECLs are entered by satisfying the Initiating Condition (IC) through meeting an Emergency Action Level (EAL) of the IC provided in this section of the Annex. The ECLs are escalated from least severe to most severe according to relative threat to the health and safety of the public and emergency workers. Depending on the severity of an event, prior to returning to a standard day-to-day organization, a state or phase called Recovery may be entered to provide dedicated resources and organization in support of restoration and communication activities following the termination of the emergency.

Unusual Event (UE): 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.

Alert: 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.

Site Area Emergency (SAE): Events are in progress or have occurred which involve an 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.

General Emergency (GE): 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 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear

  • Recovery: Recovery can be considered as a phase of the emergency and is entered by meeting emergency termination criteria provided in EP-CE-111 Emergency Classification and Protective Action Recommendations.

Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The Emergency Classification Levels, in ascending order of severity, are:

  • Unusual Event (UE)
  • Alert
  • Site Area Emergency (SAE)
  • General Emergency (GE)

Initiating Condition (IC): An event or condition that aligns with the definition of one of the four Emergency Classification Levels by virtue of the potential or actual effects or consequences.

Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given Emergency Classification Level.

An emergency is classified by assessing plant conditions and comparing abnormal conditions to ICs and EALs, based on the designated Operational Condition (MODE). Modes 1 through 5 are based on Reactor Mode Switch Position and average reactor coolant temperature.

"Defueled" Mode was established for classification purposes under NEI 99-01 to reflect conditions where all fuel has been removed from the Reactor Pressure Vessel.

OPERATING MODES REACTOR MODE SWITCH POSITION TEMP (1) Power Operation: Run N/A (2) Startup: Refuel (a) or Startup/Hot Standby N/A (3) Hot Shutdown (a): Shutdown > 212° F (4) Cold Shutdown (a): Shutdown :s; 212° F (5) Refueling (b): Shutdown or Refuel N/A (D) Defueled: All reactor fuel removed from reactor pressure vessel (full core off load during refueling or extended outage).

(a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

Hot Matrix - applies in modes (1 ), (2), and (3)

Cold Matrix - applies in modes (4), (5), and (D)

July 2019 JAF 1-2 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear Individuals responsible for the classification of events will refer to the IC's and EALs on the

  • matrix of the appropriate station Standardized Emergency Plan Annex (this document). This matrix will contain ICs, EALs, Mode Applicability Designators, appropriate EAL numbering system, and additional guidance necessary to classify events. It may be provided as a user aid.

The matrix is set up in six Recognition Categories. The first is designated as "R" and relates to Abnormal Radiological Conditions / Abnormal Radiological Effluent Releases. The second is designated as "F" and relates to Fission Product Barrier Degradation. The third is designated as "M" and relates to hot condition System Malfunctions. The fourth is designated as "C" and relates to Cold Shutdown / Refueling System Malfunctions. The fifth is designated as "H" and relates to Hazards and Other Conditions Affecting Plant Safety. The sixth is designated "E-H" and relates to ISFSI Malfunctions.

The matrix is designed to provide an evaluation of the Initiating Conditions from the worst conditions (General Emergencies) on the left to the relatively less severe conditions on the right (Unusual Events). Evaluating conditions from left to right will reduce the possibility that an event will be under classified. All Recognition Categories should be reviewed for applicability prior to classification.

  • The Initiating Conditions are coded with a two letter and one number code. The first letter is the Recognition Category designator, the second letter is the classification Level, "U" for (Notification Of) Unusual Event, "A" for Alert, "S" for Site Area Emergency and "G" for General Emergency. The EAL number is a sequential number for that Recognition Category series. All ICs that are describing the severity of a common condition (series) will have the same number.

The EAL number may then be used to reference a corresponding page(s), which provides the basis information pertaining to the IC:

  • Mode Applicability
  • Basis Classification is not to be made without referencing, comparing and satisfying the specified Emergency Action Levels.

A list of definitions is provided as part of this document (Section 1.7) for terms having specific meaning to the EALs. Site specific definitions are provided for terms with the intent to be used for a particular IC/EAL and may not be applicable to other uses of that term at other sites, the Emergency Plan or procedures.

References are also included to documents that were used to develop the EALs .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear.

References to the Emergency Director means the person in Command and Control as defined

  • in the Emergency Plan. Classification of emergencies is a non-delegable responsibility of Command and Control for the onsite facilities with responsibility assigned to the Shift Emergency Director (Control Room Shift Manager) or the Site Emergency Director (Technical Support Center). Classification of emergencies remains the responsibility of the Shift Emergency Director until Command and Control is transferred to the Site Emergency Director (Technical Support Center).

Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL is IMMINENT. If, in the judgment of the Emergency Director, an IMMINENT situation is at hand, the classification should be made as if the EAL has been exceeded. While this is particularly prudent at the higher ECL (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all ECLs.

1.2 Classification, Instrumentation and Transient Events All classifications are to be based upon valid indications, reports or conditions. Indications, reports or conditions are considered valid when they are verified by (1) an instrument channel check, or (2) indications on related or redundant indications, or (3) by direct observation by plant personnel, such that doubt related to the indication's operability, the condition's existence, or the report's accuracy is removed. Implicit in this is the need for timely assessment.

  • Indications used for monitoring and evaluation of plant conditions include the normally used instrumentation, backup or redundant instrumentation, and the use of other parameters that provide information that supports determination if an EAL has been reached. When an EAL refers to a specific instrument or indication that is determined to be inaccurate or unavailable, then alternate indications shall be used to monitor the specified condition.

During an event that results in changing parameters trending towards an EAL classification, and instrumentation that was available to monitor this parameter becomes unavailable or the parameter goes off scale, the parameter should be assumed to have been exceeded consistent with the trend and the classification made if there are no other direct or indirect means available to determine if the EAL has not been exceeded.

The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the EAL to be exceeded (i.e., this is the time that the EAL information is first available) .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear -

Planned ~volutions involve preplanning to address the limitations imposed by the condition, the

  • performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition in accordance with the specific requirements of the site's Technical Specifications. Activities which cause the site to operate beyond that allowed by the site's Technical Specifications, planned or unplanned, may result in an EAL being met or exceeded. Planned evolutions to test, manipulate, repair, perform maintenance or modifications to systems and equipment that result in an EAL being met or exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the specific operating license.

However, these conditions may be subject to the reporting requirements of 10 CFR 50.72.

When two or more EALs are determined, declaration will be made on the highest classification level.

Concerning ECL Downgrading, Exelon Nuclear policy is that ECLs shall not be downgraded to a lower classification. Once declared, the event shall remain in effect until no classification is warranted or until such time as conditions warrant classification to Recovery.

There may be cases in which *a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review), and the condition no longer exists. In these cases, an emergency should not be declared. Reporting requirements of 10 CFR 50.72 are applicable, the guidance of NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73 and the Reportability Reference Manual, should be applied .

  • 1.3 Mode Applicability The plant-operating mode that existed at the time that the event occurred, prior to any protective system or operator action initiated in response to the condition, is compared to the mode applicability of the EALs. If an event occurs, and a lower or higher plant-operating mode is reached before the emergency classification can be made, the declaration shall be based on the mode that existed at the time the event occurred.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that have Cold Shutdown or Refueling for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the Fission Product Barrier Matrix EALs are applicable only to events that initiate in Hot Shutdown or higher.

If there is a change in Mode following an event declaration, any subsequent events involving EALs outside of the current declaration escalation path will be evaluated on the Mode of the plant at the time the subsequent events occur.

1.4 Emergency Director Judgment Emergency Director (ED) Judgment EALs are provided in the Hazards and Other Condition Affecting Plant Safety section and on the Fission Product Barrier (FPB) Matrix. Both of the ED Judgment EALs have specific criteria for when they should be applied .

James A. FitzPatrick Nuclear_ Power Plant Annex Exelon Nuclear The Hazards Section ED Judgment EALs are intended to address unanticipated conditions

  • which are not addressed explicitly by other EALs but warrant declaration of an emergency because conditions exist which are believed by the ED to fall under specific emergency classifications (UE, Alert, SAE or GE).

The FPS Matrix ED Judgment EALs are intended to include unanticipated conditions, which are not addressed explicitly by any of the other FPS threshold values, but warrant determination because conditions exist that fall under the broader definition for a significant Loss or Potential Loss of the fission product barrier (equal to or greater than the defined FPS threshold values).

1.5 Fission Product Barrier (FPB) Threshold A fission product barrier threshold is a pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

FPS thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary FPBs are:

Fuel Clad (FC)

Reactor Coolant System (RCS)

Containment (CT)

  • Upon determination that one or more FPS thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the FPS IC/EAL criteria to determine the appropriate ECL.

In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/

Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g.,

spent fuel pool accidents, design containment leakage following a LOCA, etc.).

1.6 Fission Product Barrier Restoration Fission Product Barriers are not treated the same as EAL threshold values. Conditions warranting declaration of the loss or potential loss of a FPS may occur resulting in a specific classification. The condition that caused the loss or potential loss declaration could be rectified as the result of Operator action, automatic actions, or designed plant response. Barriers will be considered re-established when there are direct verifiable indications (containment penetration or open valve has been isolated, coolant sample results, etc) that the barrier has been restored and is capable of mitigating future events .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear The reestablishment of a FPB does not alter or lower the existing classification. Termination

  • 1.7 and entry into Recovery phase is still required for exiting the present classification. However the reestablishment of the barrier should be considered in determining future classifications should plant conditions or events change.

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

CONTAINMENT CLOSURE: The procedurally defined actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

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.

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fire. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be

  • met by the station.
  • HOSTILE ACTION: An act toward a Nuclear Power Plant (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).

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.

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.

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.

NORMAL LEVELS: As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear 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 .

PROJECTILE: An object directed toward a Nuclear Power Plant (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.

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.

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.

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.

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 .

  • VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train 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 SAFETY SYSTEM train .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear EMERGENCY ACTION LEVEL TECHNICAL BASIS PAGE INDEX

  • EAL General RG1 Pg.

2-28 RG2 2-38 EAL Site Area RS1 RS2 2-39 Pg.

2-30 EAL RA1 RA2 RA3 Alert Pg.

2-32 2-40 2-42 Unusual EAL RU1 RU2 RU3 Event Pg.

2-35 2-43 2-45 FG1 2-49 FS1 2-50 FA1 2-51 Fuel Clad RCS Containment FC1 2-52 FC2 2-53 RC2 2-57 CT2 2-65 RC3 2-59 CT3 2-66 RC4 2-60 FC5 2-55 RCS 2-63 CT5 2-68 CT6 2-69 FC? 2-56 RC? 2-64 CT? 2-72 MG1 2-73 MS1 2-75 MA1 2-77 MU1 2-79 MG2 2-80 MS2 2-82 MS3 2-83 MA3 2-85 MU3 2-87 MA4 2-90 MU4 2-93 MAS 2-95

  • CA1 CA2 2-102 2-106 MU6 MU?

CU1 CU3 2-98 2-100 2-104 2-109 CU4 2-111 CA5 2-113 CU5 2-115 CG6 2-117 CS6 2-121 CA6 2-125 CU6 2-127 HS1 2-130 HA1 2-132 HU1 2-135 HS2 2-137 HA2 2-139 HU3 2-140 HU4 2-144 HAS 2-147 HU6 2-150 HG? 2-153 HS7 2-154 HA? 2-155 HU? 2-156 E-HU1 2-157

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear I

Abnormal Rad Levels/ Radiological Effluents RG1 II][gj@l@lffil[g RS1 II][gj@l@lffil[g RA1 II][gj@l@lffil[g RU1 II][gj@l@lffil[g Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite Release of gaseous or liquid radioactivity greater than 2 times than 1000 mRem TEOE or 5000 mRem thyroid COE. than 100 mRem TEOE or 500 mRem thyroid COE. dose greater than 10 mRem TEOE or 50 mRem thyroid COE. the OOCM limits for 60 minutes or longer.

Emergency Action Level (EAL): Emergency Action Level (EAL): Emergency Action Level (EAL): Emergency Action Level (EAL):

Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has been Notes:

The Emergency Director should declare the event promptly upon detem,ining that the applicable lime has been Notes:

The Emergency Director should declare the event promptly upon detem,ining that the applicable lime has Notes:

The Emergency Director should declare the event promptly upon detem,ining that the applicable lime has

. exceeded, or will likely be exceeded.

If an ongoing release is detected and the release start lime is unknown, assume that the release duration has

. 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

. been exceeded, or will likely be exceeded.

lf an ongoing release is detected and the release start time is unknown, assume that the release duration 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

. exceeded 15 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the

. exceeded 15 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If

. exceeded 60 minutes.

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

C Cl>

. purposes.

The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency

. purposes.

The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency

. classification purposes.

The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency classification purposes.

classification assessments until the results from a dose 1. Reading on the Liquid Radwaste Effluent Monitor (17RM-iE classification assessments until the results from a dose classification assessments until the results from a dose 350) > 2 times hi-hi trip for;, 60 minutes.

w assessment using actual meteorology are available. assessment using actual meteorology are available. assessment using actual meteorology are available.

'iu OR

1. Readings on ANY Table R1 Effluent Monitor> Table R1 value 0 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for.=:, 15 minutes. 2. Readings on ANY Table R1 Effluent Monitor> Table R1 value
  • 5, for?, 15 minutes. for.=:, 15 minutes.

0 OR for.=:, 60 minutes.

OR 0 OR 2. Dose assessment using actual meteorology indicates doses at OR

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

2. Dose assessment using actual meteorology indicates doses at 3. Confinned sample analyses for gaseous or liquid releases l'CS or beyond the site boundary of EITHER: a. > 10 mRem TEOE D:: or beyond the site boundary of EITHER: indicate concentrations or release rates
a. > 1000 mRem TEDE OR
a. > 100 mRem TEOE > 2 times OOCM Limit with a release duration OR b. > 50 mRem COE Thyroid of a:, 60 minutes.
b. > 5000 mRem CDE Thyroid OR OR OR b. > 500 mRem COE Thyroid 3. Analysis of a liquid effluent sample indicates a concentration
3. Field survey results at or beyond the site boundary indicate or release rate that would result in doses greater than OR EITHER of the following at or beyond the site boundary.

EITHER:

a. Gamma (closed window) dose rates 3. Field survey results at or beyond the site boundary indicate a. 1 O mRem TEOE for 60 minutes of exposure.

EITHER: OR

> 1000 mR/hr are expected to continue for.=:, 60 minutes. b. 50 mRem COE Thyroid for 60 minutes

a. Gamma (closed window) dose rates of exposure.

OR > 100 mR/hr are expected to continue for=:: 60 minutes. OR

b. Analyses of field survey samples indicate
4. Reid survey results at or beyond the site boundary indicate

> 5000 mRem CDE Thyroid for 60 minutes OR EITHER:

of inhalation. a. Gamma (closed window) dose rates

b. Analyses of field survey samples indicate

> 500 mRem CDE Thyroid for 60 minutes > 10 mR/hr are expected to continue of inhalation. for a:, 60 minutes.

OR

b. Analyses of field survey samples indicate

> 50 mRem COE Thyroid for 60 minutes of inhalation.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling 0-0efueled July 2019 HOT MATRIX JAF 2-1 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex HOT MATRIX

  • HOT MATRIX
  • Exelon Nuclear Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency Site Area Emergency Alert Unusual Event Stack 7880 mR/hr 788 mRlhr 78.8 mRlhr 0.451 mR/hr (High Range Monitor)

Rx Bldg Exh NIA NIA NIA 9.50E+05 cpm (Low Range Monitor)

Turb Bldg Exh 2.44 mR/hr 0.244 mR/hr NIA 6.72E+05 cpm (Low Range Monitor)

Radw Bldg Exh 4.74 mRlhr 0.474 mR/hr NIA NIA Refuel Floor Exh NIA NIA NIA 9.28E+05 cpm (Low Range Monitor)

July 2019 HOT MATRIX JAF 2-2 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Abnormal Rad Levels/ Radiological Effluents RG2 RS2 RA2 RU2 Spent fuel pool level cannot be restored to at least 1.00 foot Spent fuel pool level at 1.00 foot. Significant lowering of water level above, or damage to, UNPLANNED loss of water level above irradiated fuel.

for 60 minutes or longer. irradiated fuel.

Emergency Action Level {EAL): Emergency Action Level {EAL):

Emergency Action Levels {EAL}: Emergency Action Level {EAL):

Lowering of spent fuel pool level to 1.00 foot as indicated

1. a. UNPLANNED water level drop in the REFUELING Note: The Emergency Director should declare the event on 19Ll-60A or 19Ll-60B.

PATHWAY as indicated by ANY of the following:

promptly upon determining that the applicable time 1. Uncovery of irradiated fuel in the REFUELING has been exceeded, or will likely be exceeded. PATHWAY.

  • Inability to restore and maintain Spent Fuel Pool water level> low water level alarm.

OR Spent fuel pool level cannot be restored to at least OR 1.00 foot as indicated on19Ll-60A or 19Ll-60B 2. Damage to irradiated fuel resulting in a release of for?. 60 minutes. radioactivity from the fuel as indicated by ANY

  • Indication or report of a drop in water level Table R2 Refuel Floor Radiation Monitors Table R2 Radiation Monitor Alarm. in the REFUELING PATHWAY.

OR AND

3. Lowering of spent fuel pool level to 11.00 feet as b. UNPLANNED Area Radiation Monitor reading rise

.l!l

  • 17RI S-456A or B Refuel Floor Exhaust C

a, ii:

w RA3 RU3 cu Radiation levels that impede access to equipment necessary Reactor coolant activity greater than Technical Cl for normal plant operations, cooldown or shutdown. Specification allowable limits.

0 Table R4 0 Areas with Entry Related Mode Applicability Emergency Action Level {EAL}: Emergency Action Level {EAL}:

Table R3

=c Areas Requiring Continuous Occupancy Entry Note: If the equipment in the room or area listed in Table 0:: Related R4 was already inoperable, or out of service, before 1 Offgas radiation?. hi-hi alarm

  • Main Control Room - (by survey) Area the event occurred, then no emergency classification Mode
  • Central Alarm Station - (by survey) Applicability is warranted. OR
1. Dose rate> 15 mR/hr in ANY of the areas in Table R3. 2. Specific coolant activity> 2.0 µCi/gm 1-131 dose
  • Reactor Building East Crescent equivalent.

OR

  • Reactor Building West Crescent
  • Reactor Building 272' Elevation 2. UNPLANNED event results in radiation levels that Mode 3, 4, prohibit or significantly impede access to ANY of the
  • Reactor Building 300' Elevation and 5 areas in Table R4.
  • Relay Room
  • North Cable Room Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defue\ed July 2019 HOT MATRIX JAF 2-3 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity> 300 uCifgm 1-131 dose None None

1. RCS Activity None None None equivalent.
2. RPV water level E!!!.!1Q! be restored and 1. RPV water level £fil!!!Q! be restored and
1. SAOG entry required maintained> o inches (TAF). maintained> o inches (TAF). SAOG entry required
2. RPV Water Level None None OR OR
3. RPV water level cannot be determined. 2. RPV water level cannot be determined.
3. Primary Containment pressure> 56 psig.

OR

1. a. Primary Containment pressure 1. UNPLANNED rapid drop in Primary 4. a. Primary Containment hydrogen
3. Primary > 2.7 psig. Containment pressure following Primary concentration.:: 6%.

Containment AND Containment pressure rise. AND None None None OR Pressure/ b. Primary Containment pressure rise is b. Primary Containment oxygen concentration Conditions due to RCS leakage. 2. Primary Containment pressure response not .::_5%.

consistent with LOCA conditions. OR

5. Heat Capacity Temperature Limit (HCTL)

(EOP-11) exceeded.

3. UNISOLABLE primary system leakage that results in EITHER of the following:
1. UNISOLABLE Main Steam Line (MSL),

HPCl, RWCU, RCIC, or Feedwater line a. Secondary Containment area temperature break. > EOP-5 Maximum Nonna! Operating

4. RCS Leak Rate None None Limit. None None OR
2. Emergency RPV Depressurization is OR required. b. Secondary Containment area radiation

> EOP-5 Maximum Nonna! Operating Limit.

5. Primary Drywell radiation monitor reading Drywell radiation monitor reading Drywell radiation monitor reading Containment None None None

> 1.8E+03 R/hr (1800 R/hr}. > 63 R/hr. > 1.8E+04 R/hr (18,000 R/hr).

Radiation

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal.

OR

2. Intentional Primary Containment venting or purging per EOPs or SAOGs due to accident conditions.
6. Primary OR Containment None None None None None Isolation Failure 3. UNISOLABLE primary system leakage that results in EITHER of the following:
a. Secondary Containment area temperature

> EOP-5 Maximum Safe Operating Limit.

OR

b. Secondary Containment area radiation

> EOP-5 Maximum Safe Operating Limit.

7. Emergency 1. Any Condition in the opinion of the 2. Any Condition in the opinion or the 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the 1. Any Condition in the opinion of the 2. Any Condition in the opinion of the Director Emergency Director that indicates Loss Emergency Director that indicates Emergency Director that indicates Loss of Emergency Director that indicates Potential Emergency Director that Indicates Loss of the Emergency Director that indicates Potential Judgment of the Fuel Clad Barrier. Potential Loss of the Fuel Clad Barrier. the RCS Barrier. Loss of the RCS Barrier. Containment Barrier. Loss of the Containment Barrier.

Modes: 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D- Defueled July 2019 HOT MATRIX JAF 2-4 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear System Malfunction MG1 I!]~[] MS1 I!]~[] MA1 I!]~[] MU1 I!]~[]

Prolonged loss of all offsite and all onsite AC power to Loss of all offsite and onsite AC power to emergency Loss of all but one AC power source to emergency buses for Loss of all offsite AC power capability to emergency buses emergency buses. buses for 15 minutes or longer. 15 minutes or longer. for 15 minutes or longer.

Emergencl£ Action Level (EAL}: Emergencl£ Action Level (EAL}: Emergencl£ Action Level (EAL}: Emergencl1 Action Level (EAL}:

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

Gi

~

a..

1. Loss of ALL offsite and onsite AC power to 4160 V emergency buses 10500 and 10600. 1. Loss of ALL offsite and onsite AC power to 4160 V 1. AC power capability to 4160 V emergency buses 10500 Loss of ALL offsite AC power capability 4160 V emergency and 10600 reduced to only one of the following power 0

...<I: 2.

AND EITHER of the following:

emergency buses 10500 and 10600.

AND ..

sources for.::_ 15 minutes.

Reserve Station Transformer T-2 buses 10500 and 10600 for.::_ 15-minutes.

Reserve Station TransformerT-2 0

U)

U) 0

a. Restoration of at least one 4160 V emergency bus 10500 or 10600 in< 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.
2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 in< 15 minutes from . Reserve Station Transformer T-3 Station Service Transformer T-4 (While

. Reserve Station Transformer T-3 Station Service Transformer T-4 (While backfeeding from Main Transformer)

OR the time of loss of both offsite and onsite AC power.

-I

b. RPV water level cannot be restored and maintained

> -19 inches (MSCRWL).

.. backfeeding from Main Transformer)

EDGA

.. EOG B EDGC AND

. EOG D Main Generator via T-4

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

MG2 I!]~[] MS2 I!]~[]

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

longer.

Emergencl£ Action Level (EAL}:

Emergencl£ Action Level (EAL}:

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

3: been exceeded, or will likely be exceeded.

0 a.. 1. Loss of ALL offsite and onsite AC power to 4160 V Voltage is < 105 VDC on Vital DC buses 71 BCB-2A 0 and 71 BCB-28 for.::_ 15 minutes.

C 0

emergency buses 10500 and 10600.

AND U) 2. Voltage is < 105 VDC on Vital DC buses 71 BCB-2A and U) 0 71BCB-2B.

-I AND

3. All AC and Vital DC power sources in EALs #1 and #2 have been lost for.::_ 15 minutes.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 HOT MATRIX JAF 2-5 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear I

System Malfunction MS3 MA3 MU3 Inability to shutdown the reactor causing a challenge to RPV Automatic or manual scram fails to shutdown the Automatic or manual scram fails to shutdown the reactor.

water level or RCS heat removal. reactor. and subsequent manual actions taken at the Emergency Action Level CEALl:

Reactor Control Console are not successful in shutting Emergency Action Level (EAL):

down the reactor. Note: A manual action is any operator action, or set of

1. Automatic scram did not shutdown the reactor as actions, which causes the control rods to be rapidly Emergency Action Level (EAU:

indicated by Reactor Power:::. 2.5%. inserted into the core. This action does not include Note: A manual action is any operator action, or set of manually driving in control rods or implementation of AND actions, which causes the control rods to be rapidly boron injection strategies.

2. ALL manual I ARI actions to shutdown the reactor have inserted into the core. This action does not include 1 a. Automatic scram did not shutdown the reactor as been unsuccessful as indicated by Reactor Power manually driving in control rods or implementation of
  • indicated by Reactor Power:::. 2.5%.
,2.5%. boron injection strategies.

AND AND 1. Automatic or manual scram did not shutdown the

.b. Subsequent manual I ARI action taken at the reactor as indicated by Reactor Power:::. 2.5%.

3. EITHER of the following conditions exist: Reactor Control Console is successful in shutting AND down the reactor as indicated by
  • RPV water level cannot be restored and maintained Reactor Power< 2.5%.

> -19 inches (MSCRWL). 2. Manual / ARI actions taken at the Reactor Control Console are not successful in shutting down the reactor OR OR as indicated by Reactor Power:::. 2.5%. 2. a. Manual scram did not shutdown the reactor as

  • Heat Capacity Temperature Limit (HCTL) indicated by Reactor Power:::. 2.5%.

(EOP-11) exceeded. AND

b. EITHER of the following:
1. Subsequent manual I ARI action taken at the Reactor Control Console is successful in shutting down the reactor as indicated by Reactor Power< 2.5%.

OR

2. Subsequent automatic scram / ARI is successful in shutting down the reactor as indicated by Reactor Power< 2.5%.

MA4 MU4 II) UNPLANNED loss of Control Room indications for 15 UNPLANNED loss of Control Room indications for 15 C

minutes or longer with a significant transient in progress. minutes or longer.

~ra Table M2 Significant Transients Emergency Action Level (EAU: Emergency Action Level (EAU:

Table M1 Control Room Parameters "C

-=E

.. Reactor Power

  • Auto/Manual runback > 25% thermal reactor power Note: The Emergency Director should declare the event promptly upon determining that the applicable time Note: The Emergency Director should declare the event promptly upon determining that the applicable time RPV Water Level has been exceeded, or will likely be exceeded. has been exceeded. or will likely be exceeded.

0 RPV Pressure 0

  • Electric load rejection > 25% full electric 0: Primary Containment Pressure load 1. UNPLANNED event results in the inability to monitor UNPLANNED event results in the inability to monitor ANY

~ . Torus Level ANY Table M1 parameter from within the Control Room Table M1 parameter from within the Control Room

  • Reactor Scram for:::. 15 minutes. for:::. 15 minutes.

C Torus Temperature 0

  • EGGS injection AND 0
  • Thermal Power oscillations > 10% (peak to 2. ANY Table M2 transient in progress.

peak)

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuehng D-Defueled July 2019 HOT MATRIX JAF 2-6 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear System Malfunction MA5 l!J[gjrn Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.

Emergency Action Level fEAL):

Note:

. This EAL is only applicable to SAFETY SYSTEMs

. having two (2) or more trains.

If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred,

. then this emergency classification is not warranted.

If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not

. warranted.

If a hazardous event occurs and it is determined that the conditions of MA5 are not met, then assess the event via HU3, HU4, or HU6.

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

Seismic event (earthquake)

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

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

E

.l!l II)

AND en

~

.S!ra C.

EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical en Specifications for the current operating mode.

£!

(J

~ra

...ra "O

. OR Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the N

ra current operating mode.

r:

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-7 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear System Malfunction MU6 [TI~~

RCS leakage for 15 minutes or longer.

Emergenc~ Action Level {EAL):

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

1. RCS unidentified or pressure boundary leakage in the Drywell > 1O gpm for::: 15 minutes.

OR

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

OR

3. Leakage from the RCS to a location outside the Drywell > 25 gpm for::. 15 minutes.

MU7 [TI~~

Table M3 Communications Capability Loss of all onsite or offsite communication capabilities.

System Onsite Offsite NRC Em erg enc~ Action Level {EAL):

Page/Party System X 1. Loss of ALL Table M3 onsite communication (Gaitronics) capabilities affecting the ability to perform routine Control Room/Portable X operations.

Radio OR

2. Loss of ALL Table M3 offsite communication Plant Telephones (all X X X 1/) capabilities affecting the ability to perform offsite C VOiP, switched, non- notifications.

0 switched)

OR

~.., Installed Out-of-Plant X X X 3. Loss of ALL Table M3 NRG communication "i: Cellular Phones capabilities affecting the ability to perform NRG E Plant Satellite Phones X X notifications.

E (Installed in CR and 0

CJ deployable)

RECS X Dedicated Phone Lines X X

(ENS)

HPN and FTS 2001 X X

Modes: 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-8 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HS1 BJ~~ @I [fil [g HA1 BJ~~ @l[fil[g HU1 BJ~~ @I [fil [g HOSTILE ACTION within the PROTECTED AREA. HOSTILE ACTION within the OWNER CONTROLLED Confirmed SECURITY CONDITION or threat.

AREA or airborne attack threat within 30 minutes.

Emergenc~ Action Level {EAL): Emergenc~ Action Level {EAL):

Emergenc~ Action Level {EAL}:

A notification from the Security Supervisor that a HOSTILE 1. Notification of a credible security threat directed at the C:

ACTION is occurring or has occurred within the 1. A validated notification from NRG from an aircraft site as determined per SY-AA-101-132, Security

8

<J PROTECTED AREA. attack threat< 30 minutes of the site. Assessment and Response to Unusual Activities.

ct OR OR

2. A validated notification from the NRG providing
E Cl)
2. Notification by the Security Supervisor that a information of an aircraft threat.

0 HOSTILE ACTION is occurring or has occurred

c OR within the OWNER CONTROLLED AREA.
3. Notification by the Security Supervisor of a SECURITY CONDITION that does not involve a HOSTILE ACTION.

HS2 BJ~~@l[fil[g HA2 BJ~~@l[fil[g Inability to control a key safety function from outside the Control Room evacuation resulting in transfer of plant control Table H1 Safety Functions Control Room. to alternate locations.

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

Emergenc~ Action Level {EAL): Emergenc~ Action Level {EAL):

0

~ . RPV Water Level (ability to cool the core)

Note: The Emergency Director should declare the event promptly upon determining that the applicable time has A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate C: locations per AOP-43, Plant Shutdown from Outside the 0 been exceeded, or will likely be exceeded.

(.) RCS Heat Removal (ability to maintain heat sink) Control Room.

C:

ns 1. A Control Room evacuation has resulted in plant control ii:

0

.S!Cl) being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the Control Room .

C: AND

...ns I- 2. Control of ANY Table H1 key safety function is not reestablished in < 30 minutes.

Modes: 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 HOT MATRIX JAF 2-9 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HU3 Table H2 Areas

. Reactor Building (when inerted the Drywell is exempt)

FIRE potentially degrading the level of safety of the plant.

Emergency Action Level CEALl:

. Control Room I Relay Room I Cable Run Rooms I Cable Spreading Note:

  • The Emergency Director should declare the Room

. Electric Bays event promptly upon determining that the applicable lime has been exceeded. or will likely

. Control Room AC Equipment Room be exceeded.

.. Control Room Chiller Room Emergency Diesel Generator Building

  • Escalation of the emergency classification level would be via IC CA2 or MA5.

. Battery Rooms I Battery Room Corridor only

1. A FIRE in ANY Table H2 area is not extinguished in < 15 minutes of ANY of the following FIRE detection

. indications:

.. RHRSW / ESW Pump Rooms Cable Tunnels Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MSIV /

  • Field verification of a single fire alarm ADS) 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 Fl RE is not verified in < 30 minutes of alarm receipt.

OR

3. A FIRE within the plant PROTECTED AREA not extinguished in < 60 minutes of the initial report, alarm or indication.

OR

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

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 HOT MATRIX JAF 2-10 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HU4 ITI lli@l@J@l ITl Seismic event greater than QBE levels.

Emergenc11 Action Level {EAL}:

Note:

. For emergency classification if EAL# 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Emergency Director in~ 15 minutes of the event.

. Escalation of the emergency classification level would be via IC CA2 or MA5.

1. Seismic event> Operating Basis Earthquake (OBE) as determined by seismic monitoring system in accordance with AOP-14 Earthquake.

OR

2. When Seismic Monitoring Equipment is not available:
a. Control Room personnel feel an actual or potential Q) seismic event.

l'CS AND C" b. ANY one of the following confirmed in~ 15 minutes of

.c t:: the event:

w l'CS

. 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 a manner deemed appropriate by the Shift Manager or Emergency Director.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-11 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety Table H3 HAS @]@!~

Areas with Entry Related Mode Applicability Gaseous release impeding access to equipment Entry Related necessary for normal plant operations, cooldown or Area Mode shutdown.

U)

(II Aoolicability Emergency Action Level (EAL):

(!)

u "i<

.. Reactor Building East Crescent Note: If the equipment in the listed room or area was already inoperable, or out of service, before the 0

I- .. Reactor Building West Crescent Reactor Building 272' Elevation Mode 3, 4, event occurred, then no emergency classification is warranted.

Reactor Building 300' Elevation Relay Room and 5 1. Release of a toxic, corrosive, asphyxiant or North Cable Room flammable gas in a Table H3 area.

AND

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

HUS II]~[fil@l[fil!g Hazardous Event Emergency Action Level (EAL}:

Note:

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

. Escalation of the emergency classification level would w

C GI

> OR be via IC CA2 or MA5.

1. Tornado strike within the PROTECTED AREA.

U)

I 2. Internal room or area flooding of a magnitude sufficient to 0 require manual or automatic electrical isolation of a "C

(II SAFETY SYSTEM component required by Technical N Specifications for the current operating mode.

(II

c OR
3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Intake Water Level > 255 feet.

OR

6. ESW intake bay water level ~ 237 feet.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuellng D-Defueled July 2019 HOT MATRIX JAF 2-12 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY HOT MATRIX SITE AREA EMERGENCY

  • ALERT HOT MATRIX UNUSUAL EVENT
  • Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HG7 HS7 HA7 HU7 Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL Emergency Director warrant declaration of a SITE AREA Emergency Director warrant de_claration of an ALERT. Emergency Director warrant declaration of an UNUSUAL EMERGENCY. EMERGENCY. EVENT.

Emergency Action Level (EAU:

Emergency Action Level (EAL): Emergency Action Level (EAL): Emergency Action Level (EAL):

Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director, indicate that events are in progress or Other conditions exist which in the judgment of the Emergency Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve an actual or potential Director indicate that events are in progress or have occurred have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of substantial degradation of the level of safety of the plant or a which indicate a potential degradation of the level of safety of core degradation or melting with potential for loss of plant functions needed for protection of the public or HOSTILE security event that involves probable life threatening risk to the plant or indicate a security threat to facility protection has containment integrity or HOSTILE ACTION that results in an ACTION that results in intentional damage or malicious acts, site personnel or damage to site equipment because of been initiated. No releases of radioactive material requiring actual loss of physical control of the facility. Releases can be (1) toward site personnel or equipment that could lead to the HOSTILE ACTION. Any releases are expected to be limited offsite response or monitoring are expected unless further reasonably expected to exceed EPA Protective Action likely failure of or, (2) that prevent effective access to to small fractions of the EPA Protective Action Guideline degradation of safety systems occurs.

Guideline exposure levels offsite for more than the immediate equipment needed for the protection of the public. Any exposure levels.

C:

G) site area . releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

E Cl "t:,

.su I!!

c 1;'

C:

G)

G)

E w

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 HOT MATRIX JAF 2-13 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex HOT MATRIX

  • HOT MATRIX
  • Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 II]~@l@l[fil[g Damage to a loaded cask CONFINEMENT BOUNDARY.

Emergencll Action Level {EAL}:

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading> ANY Table E-1 values:

Table E-1 Radiation Reading Overpack Overpack Overpack Overpack Serial Average Serial Average Surface Number Surface Dose Number Dose Rates Rates mrem/hr mrem/hr (gamma+neutron) HI- (gamma+neutron)

STORM 100S (XXX)

HI- BO on the side S/N- 600 on the side STORM 0307, 20 on the top 0308, 60 on the top 100S cij

u. 32 at the inlet 0309, S/N-15. 0310,

~ 16, 17 and outlet vent ducts 0311, 0312, 0679, 0680, 0681, HI- 100 on the side 0682, STORM 0683, 20 on the top 100s 0690, (232) 90 at the inlet 0691, and outlet 0692, S/N-vent ducts 0693, 0169, 0694, 0170, 0695, 0171 1172, 1173, HI- 1174, STORM 1175, 100S 1176, (XXX) 1177 SIN- 220 on the side 0186, 0187, 40 on the top 0188 Modes: 1 - Power Operation 2- Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019. HOT MATRIX JAF 2-14 HOT MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluents RG1 II]~@l@J@l[g RS1 II]~@l@J@l[g RA1 II]~@l@J@l[g RU1 II]~@l@J@l[g Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite dose Release of gaseous or liquid radioactivity greater than 2 than 1000 mRem TEDE or 5000 mRem thyroid COE. than 100 mRem TEDE or 500 mRem thyroid COE. greater than 10 mRem TEDE or 50 mRem thyroid COE. times the ODCM limits for 60 minutes or longer.

Emergency Action Level (EAL): Emergency Action Level IEALl: Emergency Action Level (EAL): Emergency Action Level (EAL):

Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has been Notes:

. The Emergency Director should declare the event promptly upon determining that the applicable time has been Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, Notes:

The Emergency Director should declare the event promptly upon determining that the applicable time has

. 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, or will likely be exceeded.

If an ongoing release is detected and the release start time is unknown, assume that the release duration has

. 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

. 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

. exceeded 15 minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the

. minutes.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent

. exceeded 60 minutes.

Classification based on effluent monitor readings assumes that a release path to the enyironment is effluent flow past an effluent monitor is known to have effluent flow past an effluent monitor is known to have flow past an effluent monitor is known to have stopped due to established. If the effluent flow past an effluent monitor stopped due to actions to isolate the release path, then the stopped due to actions to isolate the release path, then the actions to isolate the release path, then the effluent monitor is known to have stopped due to actions to isolate the J!!

C Cl) effluent monitor reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in .

effluent monitor reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in

. reading is no longer valid for classification purposes.

The pre-calculated effluent monitor values presented in EAL

  1. 1(Table R1) should be used for emergency classification release path, then the effluent monitor reading is no longer valid for classlfication purposes.
, EAL #1(Table R1) should be used for emergency EAL #1(Table R1) should be used for emergency assessments until the results from a dose assessment using
1. Reading on the Liquid Radwaste Effluent Monitor 5: classification assessments until the results from a dose classification assessments until the results from a dose actual meteorology are available.

w assessment using actual meteorology are available. assessment using actual meteorology are available. (1 ?RM-350) > 2 times hi-hi trip for~ 60 minutes.

iii 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for OR u 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value 1. Readings on ANY Table R1 Effluent Monitor> Table R1 value  ::. 15 minutes.

"c, for::. 15 minutes. for 2. 15 minutes. OR

2. Readings on ANY Table R1 Effluent Monitor> Table R1 0 value for 2, 60 minutes.

0 OR OR 2. Dose assessment using actual meteorology indicates doses at or OR

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

nl 2. Dose assessment using actual meteorology indicates doses at a. > 10 mRem TEDE 3. Confirmed sample analyses for gaseous or liquid releases or beyond the site boundary of EITHER:

D:: or beyond the site boundary of EITHER: OR indicate concentrations or release rates

a. > 1000 mRem TEDE
a. > 100 mRem TEDE b. > 50 mRem CDE Thyroid > 2 times ODCM Limit with a release duration DR OR of~ 60 minutes.
b. > 5000 mRem COE Thyroid OR
3. Analysis of a liquid effluent sample indicates a concentration or OR b. > 500 mRem COE Thyroid
  • release rate that would result in doses greater than EITHER of
3. Field survey results at or beyond the site boundary indicate the following at or beyond the site boundary OR EITHER: a. 1 O mRem TEDE for 60 minutes of exposure
a. Gamma (closed window) dose rates 3. Field survey results at or beyond the site boundary indicate OR EITHER:

> 1000 mR/hr are expected to continue b. 50 mRem COE Thyroid for 60 minutes for 2. 60 minutes. a. Gamma (closed window) dose rates of exposure OR > 100 mR/hr are expected to continue OR

b. Analyses of field survey samples indicate

> 5000 mRem COE Thyroid for 60 minutes of inhalation. b.

for 2, 60 minutes.

OR Analyses offield survey samples indicate

4. Field survey results at or beyond the site boundary indicate EITHER:
a. Gamma (closed window) dose rates

> 500 mRem COE Thyroid for 60 minutes > 10 mR/hr are expected to continue of inhalation. for~ 60 minutes.

OR

b. Analyses of field survey samples indicate

> 50 mRem COE Thyroid for 60 minutes of inhalation.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-15 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear Table R1 Effluent Monitor Thresholds Effluent Monitor General Emeraencv Site Area Emeraencv Alert Unusual Event Stack 7880 mR/hr 788 mR/hr 78.8 mR/hr 0.451 mR/hr (High Range Monitor)

Rx Bldg Exh N/A N/A N/A 9.50E+05 cpm (Low Range Monitor)

Turb Bldg Exh 2.44 mR/hr 0.244 mR/hr N/A 6.72E+05 cpm (Low Range Monitor)

Radw Bldg Exh 4.74 mR/hr 0.474 mR/hr N/A N/A Refuel Floor Exh N/A N/A N/A 9.28E+05 cpm (Low Range Monitor)

July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-16 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear I

Abnormal Rad Levels/ Radiological Effluents RG2 RS2 RA2 RU2 Spent fuel pool level cannot be restored to at least 1.00 foot Spent fuel pool level at 1.00 foot. Significant lowering of water level above, or damage to, UNPLANNED loss of water level above irradiated fuel.

for 60 minutes or longer. irradiated fuel.

Emergency Action Level (EAL): Emergency Action Level !EAL):

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

Lowering of spent fuel pool level to 1.00 foot as indicated Note: The Emergency Director should declare the event 1. a. UNPLANNED water level drop in the REFUELING on 19Ll-60-A or 19Ll-60B.

promptly upon determining that the applicable time PATHWAY as indicated by ANY of the following:

1. Uncovery of irradiated fuel in the REFUELING has been exceeded, or will likely be exceeded. PATHWAY.
  • Inability to restore and maintain Spent Fuel Pool water level> low water level alarm.

OR Spent fuel pool level cannot be restored to at least 1.00 foot as indicated on 19Ll-60A or 19Ll-60B 2. Damage to irradiated fuel resulting in a release of OR for~ 60 minutes. ,radioactivity from the fuel as indicated by ANY

  • Indication or report of a drop in water level in Table R2 Refuel Floor Radiation Monitors Table R2 Radiation Monitor Alarm. the REFUELING PATHWAY.

OR AND

3. Lowering of spent fuel pool level to 11.00 feet as b. UNPLANNED Area Radiation Monitor reading rise

J!l C

s

=w iii RA3 Radiation levels that impede access to equipment necessary "i5, 0

for normal plant operations, cooldown or shutdown.

Table R4 0 Emergency Action Level (EAL):

'6 Areas with Entry Related Mode Applicability Table R3 Note: If the equipment in the room or area listed in Table

~ Areas Requiring Continuous Occupancy Entry R4 was already inoperable, or out of service, before Related the event occurred, then no emergency classification

  • Main Control Room - (by survey) Area Mode is warranted.
  • Central Alarm Station - (by survey) Applicability
1. Dose rate> 15 mR/hr in ANY of the areas in Table R3.
  • Reactor Building East Crescent OR
  • Reactor Building West Crescent 2. UNPLANNED event results in radiation levels that
  • Reactor Building 272' Elevation prohibit or significantly impede access to ANY of the
  • Reactor Building 300' Elevation Mode 3, 4, areas in Table R4.

and 5

  • Relay Room
  • North Cable Room Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuelmg D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-17 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear UNUSIJAL EVENT Cold Shutdown / Refueling System Malfunctions CA1 ~[fil[g CU1 ~[fil[g Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses for for 15 minutes or longer. 15 minutes or longer.

Emergencll Action Level (EAL}: Emergencll Action Level (EAL}:

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

1. Loss of ALL offsite and onsite AC power to 4160 V 1. AC power capability to 4160 V emergency buses 10500 and emergency buses 10500 and 10600. 10600 reduced to only one of the following power sources 2.

AND Failure to restore power to at least one 4160 V emergency for?. 15 minutes.

Reserve Station Transformer T-2 bus 10500 or 10600 in < 15 minutes from the time of loss of both offsite and onsite AC power.

. Reserve Station Transformer T-3 Station Service Transformer T-4 (While backfeeding

.. from Main Transformer)

EDGA Cll EOG B EDGC EDGD 0== AND D..

u 2. ANY additional single power source failure will result in a ci:

0 UI UI 0

loss of ALL AC power to SAFETY SYSTEMS.

..I Modes: 1 - Power Operation 2 -Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refuelmg D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-18 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear UNUSUAL EVENT Cold Shutdown I Refueling System Malfunctions CA2 HJ[§ Hazardous event affecting SAFETY SYSTEM required for the current operating mode.

Emergencl£ Action Level (EAL!:

Note:

. This EAL is only applicable to SAFETY SYSTEMs

. having two (2) or more trains.

If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred, then

. this emergency classification is not warranted.

If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not

. warranted.

If a hazardous event occurs and it is determined that the conditions of CA2 are not met, then assess the event via HU3, HU4, or HU6.

E 1. a. The occurrence of ANY of the following hazardous

~>, ..

events:

Seismic event (earthquake)

Cl)

.a,

.2!

.. Internal or external flooding event High winds or tornado strike Cl) .. FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager AND

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

AND C.

EITHER of the following:

Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

OR

. Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-19 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Cold Shutdown / Refueling System Malfunctions CU3 ~[§ GI Loss of Vital DC power for 15 minutes or longer.

0 Emergency Action Level (EAL):

II.

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

Voltage is < 105 voe on required Vital DC buses 71 BCB-2A and 71 BCB-2B for:::. 15 minutes.

CU4 ~@Jig Table C1 Communication Capabilities Loss of all onsite or offsite communication capabilities.

System Onsite Offsite NRG Page/Party System Emergency Action Level (EAL):

X (Gaitronics) 1. Loss of ALL Table C1 onsite communication U)

C Control Room/Portable X capabilities affecting the ability to perform routine

8 Radio operations.
  • c"'::i Plant Telephones (all VOiP, X X X OR 0

switched, non-switched)

2. Loss of ALL Table C1 offsite communication E Installed Out-of-Plant X X X capabilities affecting the ability to perform offsite E Cellular Phones notifications.

0 0 Plant Satellite Phones X X OR (Installed in CR and deployable) 3. Loss of ALL Table C1 NRG communication capabilities affecting the ability to perform NRG RECS X notifications.

Dedicated Phone Lines X X (ENS)

HPN and FTS 2001 X X GAS ~[§ GUS ~[§ Table C2 RCS Heat-up Duration Thresholds UNPLANNED rise in RCS temperature.

Inability to maintain the plant in cold shutdown.

RCS Containment Heat-up Emergency Action Levels (EAU: Emergency Action Levels (EAU:

Status Closure Status Duration Intact Not Aoolicable 60 minutes* Note:

. The Emergency Director should declare the event Note:

The Emergency Director should declare the event promptly upon determining that the applicable time C

Established 20 minutes* promptly upon determining that the applicable time

. has been exceeded, or will likely be exceeded.

iii 1uGI Not Intact O minutes

. has been exceeded, or will likely be exceeded.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does J: Not Established not warrant classification.

limit when heat removal function is available does not warrant classification. 1. UNPLANNED rise in RCS temperature> 212 'F.

  • If an RCS heat removal system is in operation 1. UNPLANNED rise in RCS temperature> 212 'F OR within this time frame and RCS temperature is being for> Table C2 duration. 2. Loss of the following for:::_ 15 minutes.

reduced, then EAL #1 is not applicable.

OR . ALL RCS temperature indications

2. UNPLANNED RPV pressure rise> 1O psig as a result of temperature rise.

. AND ALL RPV water level indications Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-20 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear I

Cold Shutdown I Refueling System Malfunctions CGS 11][§ CS6 11][§ CAS 11][§ cus 11][§ Loss of RPV inventory affecting fuel clad integrity with Loss of RPV inventory affecting core decay heat removal Loss of RPV inventory. UNPLANNED loss of RP inventory for 15 minutes or containment challenged. capability. longer.

EmergencJl Action Level (EAL}:

Emergencll Action Level (EAL}: Emergencll Action Level (EAU: EmergencJl Action Level (EAL):

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

1. Loss of RPV inventory as indicated by 1 a. RPV water level < 0 inches (TAF) for 1. With CONTAINMENT CLOSURE not established, level < 126.5 inches. 1. UNPLANNED loss of reactor coolant results in the
. 30 minutes. RPV water level < 120.5 inches. inability to restore and maintain RPV level to above the OR AND OR procedurally established lower limit
2. With CONTAINMENT CLOSURE established, RPV 2. a. RPV water level cannot be monitored for::. 15 minutes.
b. ANY Table C4 Containment Challenge Indication. for::. 15 minutes-.- -

water level < 0 inches (TAF). OR OR OR AND

2. a. RPV water level cannot be monitored 2. a. RPV water level £fil!!1Q! be monitored.
3. a. RPV water level cannot be monitored for for::. 30 minutes-.- - b. Loss of RPV inventory per Table C3 indications.
. 30 minutes. AND

~ AND AND b . Loss of RPV inventory per Table C3 indications.

.s b. Core uncovery is indicated by ANY of the C

GI

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

Table C3 indication of a sufficient magnitude following:

Table C3 indication of a sufficient magnitude GI Cl ns .

to indicate core uncovery.

OR 1BRIA-052-30 Refuel Floor West (EPIC A- .

to indicate core uncovery.

OR 1BRIA-052-30 Refuel Floor West (EPIC A-ns 1247) Rad monitor::. 3 R/hr. 1247) Rad monitor::. 3 R/hr.

GI

..J AND ti) c. ANY Table C4 Containment Challenge Indication.

(J 0::

. Table C3 Indications of RCS Leakage UNPLANNED Drywell equipment drain sump level Table C4 Containment Challenge Indications rise*

UNPLANNED Drywell floor drain sump level rise*

. Primary Containment Hydrogen Concentration

. UNPLANNED Reactor Building equipment sump level rise* .

. 6% and Oxygen ::. 5%

UNPLANNED rise in containment pressure

. UNPLANNED Reactor Building floor drain sump level . CONTAINMENT CLOSURE not established*

rise*

UNPLANNED Torus level rise*

. Secondary Containment area radiation > ANY Maximum Safe Operating Limit (EOP-5)

. UNPLANNED RPV make up rate rise*

. Observation of leakage or inventory loss

  • if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not
  • Rise in level is attributed to a loss of RPV inventory required.

Modes: 1 - Power Operation 2- Startup 3- Hot Shutdown 4 - Cold Shutdown 5-Refuehng D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-21 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

Ja=m=e:::s:::A::::*::::F::::itz::::::P:::at::::r::::ic::::k::::N::::u::::c::::le::::a::::r::::P::::o::::w::::e::::r::::P::::l:::an::::t::::A::::n::::n::::e::::x::::::=====C:::O::::L::::D:::::::S::::H:::U::::T:::D:::O::::W'==N=/R:::E:::F::::U::::::E=L=IN:::G"::=M':':A:::T:::R:::'l:::X:::::::========================C=O=='L=D:::S=H=U=T=D'==O=WN/REFUELING MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HS1 ITI rn @l@l lfil C9 HA1 ITI rn @l@l lfil C9 HU1 ITI rn@l !1l lfil [g HOSTILE ACTION within the PROTECTED AREA. HOSTILE ACTION within the OWNER CONTROLLED Confirmed SECURITY CONDITION or threat.

AREA or airborne attack threat within 30 minutes.

Emergenc:l Action Level (EAL): Emergenc:l Action Level (EAL):

Emergenc:l Action Level (EAL):

A notification from the Security Supervisor that a HOSTILE 1. Notification of a credible security threat directed at C:

ACTION is occurring or has occurred within the 1. A validated notification from NRC of an aircraft the site as determined per SY-AA-101-132, Security

~u PROTECTED AREA. attack threat< 30 minutes from the site. Assessment and Response to Unusual Activities.

< OR OR

~ 2. Notification by the Security Supervisor that a

2. A validated notification from the NRC providing 1ii information of an aircraft threat.

0 HOSTILE ACTION is occurring or has occurred OR

r: within the OWNER CONTROLLED AREA.
3. Notification by the Security Supervisor of a SECURITY CONDITION that does not involve a HOSTILE ACTION.

HS2 ITI rn@l@l lfil[g HA2 ITI rn@l@l lfil [g Inability to control a key safety function from outside the Control Room evacuation resulting in transfer of plant Table H1 Safety Functions Control Room. control to alternate locations.

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

Emergenc:l Action Level (EAL}: Emergenc:l Action Level (EAL):

0

.l:l . RPV Water Level (ability to cool the core)

Note: The Emergency Director should declare the event promptly upon determining that the applicable time A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate C: locations per AOP-43, Plant Shutdown from Outside the 0 has been exceeded, or will likely be exceeded.

(.) RCS Heat Removal (ability to maintain heat sink) Control Room.

C:

cu 1. A Control Room evacuation has resulted in plant control

....a:0 being transferred from the Control Room to alternate

......en locations per AOP-43, Plant Shutdown from Outside the Cll Control Room.

C: AND E 2. Control of ANY Table H1 key safety function is not I-reestablished in < 30 minutes.

Modes. 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-22 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX I

SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety Table H2 Areas HU3

. Reactor Building (when inerted the FIRE potentially degrading the level of safety of the plant.

. Drywell is exempt)

Control Room I Relay Room I Cable Run Rooms I Cable Spreading Emergency Action Level (EAL):

Note:

Room

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

.. Control Room AC Equipment Room Control Room Chiller Room Emergency Diesel Generator be exceeded.

Escalation of the emergency classification level would be via IC CA2 or MA5

. Building Battery Rooms/ Battery Room 1. A FIRE in ANY Table H2 area is not extinguished in

.. Corridor only RHRSW / ESW Pump Rooms Cable Tunnels

< 15 minutes of ANY of the following FIRE detection indications:

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

. Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MSIV /

  • Receipt of multiple (more than 1) fire alarms or indications
  • Field verification of a single fire alarm ADS)

OR

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

AND

b. The existence of a FIRE is not verified in

< 30 minutes of alarm receipt.

OR

3. A FIRE within the plant PROTECTED AREA!!Q!

extinguished in < 60 minutes of the initial report, alarm or indication.

OR

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

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5-Refueling D - Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-23 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • COLD SHUTDOWN/REFUELING MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HU4 BJ~§@l@l[g Seismic event greater than QBE levels.

Emergenc~ Action Level (EAL!:

Note:

. For emergency classification if EAL# 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the

. Emergency Director in!': 15 minutes of the event Escalation of the emergency classification level would be via IC CA2 or MA5.

1. Seismic event> Operating Basis Earthquake (OBE) as determined by seismic monitoring system in accordance with AOP-14 Earthquake.

OR

2. When Seismic Monitoring Equipment is not available:
a. Control Room personnel feel an actual or potential GI seismic event.

cu

, AND l:T

.c b. ANY one of the following confirmed in!': 15 minutes

~

cu of the event:

w

. 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 a manner deemed appropriate by the Shift Manager or Emergency Director.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-24 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HAS ~[ii~

Table H3 Gaseous release impeding access to equipment Areas with Entry Related Mode Applicability necessary for normal plant operations, cooldown or shutdown.

Entry ti)

I'll Related Emergencll Action Level (EAL}:

(!) Area Mode t.)

Note: If the equipment in the listed room or area was Aoolicabilitv

  • c already inoperable, or out of service, before the 0 event occurred, then no emergency classification I- Reactor Building East Crescent is warranted.

Reactor Building West Crescent

1. Release of a toxic, corrosive, asphyxiant or

.. Reactor Building 272' Elevation Reactor Building 300' Elevation Mode 3, 4, flammable gas in a Table H3 area.

. Relay Room North Cable Room and 5 2.

AND Entry into the room or area is prohibited or impeded HUS ITJ~@l[i]ffi]lg Hazardous Event Emergencl£ Action Level (EAL}:

Note:

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

.. . Escalation of the emergency classification level would be via IC CA2 or MA5 C:

Cll

> 1. Tornado strike within the PROTECTED AREA.

w ti) OR

I 0 2. Internal room or area flooding of a magnitude sufficient to "C

I'll require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical N

I'll Specifications for the current operating mode.

J:

OR

3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Intake Water Level > 255 feet.

OR

6. ESW intake bay water level < 237 feet.

Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5- Refueling D-Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-25 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex GENERAL EMERGENCY COLD SHUTDOWN/REFUELING MATRIX SITE AREA EMERGENCY

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear Hazards and Other conditions Affecting Plant Safety HG7 HS7 HA7 HU7 Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL Emergency Director warrant declaration of a SITE AREA Emergency Director warrant declaration of an ALERT. Emergency Director warrant declaration of an UNUSUAL EMERGENCY. EMERGENCY. EVENT.

Emergency Action Level (EAL):

Emergency Action Level (EAL}: Emergency Action Level (EAL): Emergency Action Level (EAL):

Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director, indicate that events are in progress or Other conditions exist which in the judgment of the Emergency Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve an actual or potential Director indicate that events are in progress or have occurred have occurred which involve actual or IMMINENT substantial have occurred which involve actual or likely major failures of substantial degradation of the level of safety of the plant or a which indicate a potential degradation of the level of safety of core degradation or melting with potential for loss of plant functions needed for protection of the public or HOSTILE security event that involves probable life threatening risk to the plant or indicate a security threat to facility protection has containment integrity or HOSTILE ACTION that results in an ACTION that results in intentional damage or malicious acts, site personnel or damage to site equipment because of been initiated. No releases of radioactive material requiring actual loss of physical control of the facility. Releases can be (1) toward site personnel or equipment that could lead to the HOSTILE ACTION. Any releases are expected to be limited offsite response or monitoring are expected unless further reasonably expected to exceed EPA Protective Action likely failure of or, (2) that prevent effective access to to small fractions of the EPA Protective Action Guideline degradation of safety systems occurs.

Guideline exposure levels offsite for more than the immediate equipment needed for the protection of the public. Any exposure levels.

C a,

site area. releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

E C)

"C

~

f!

c

~

C a,

e>

a, E

w Modes: 1 - Power Operation 2-Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D- Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-26 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex COLD SHUTDOWN/REFUELING MATRIX

  • ALERT COLD SHUTDOWN/REFUELING MATRIX UNUSUAL EVENT Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ISFSI Malfunction E-HU1 IIlrnl:filGJ@J[g Damage to a loaded cask CONFINEMENT BOUNDARY.

Emergencll: Action Level (EAL}:

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading > ANY Table E-1 values:

Table E-1 Radiation Reading Overpack Overpack Overpack Overpack Serial Average Serial Average Surface Number Surface Dose Number Dose Rates Rates mrem/hr mrem/hr (gamma+neutron)

HI- (gamma+neutron)

STORM 100S (XXX)

HI- 80 on the side SIN- 600 on the side STORM 0307.

20 on the top 0308. 60 on the top 100S en

u. S/N-15. 32 at the inlet 0309.

and outlet 0310.

~ 16.17 0311.

vent ducts 0312.

0679.

0680.

0681.

HI- 100 on the side 0682, STORM 0683, 20 on the top 100s 0690, (232) 90 at the inlet 0691, and outlet 0692, SIN- 0693, vent ducts 0169, 0694, 0170, 0695, 0171 1172, HI- 1173, STORM 1174, 100S 1175, (XXX) 1176, 1177 SIN- 220 on the side 0186, 0187, 40 on the top 0188 Modes: 1 - Power Operation 2 -Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled July 2019 COLD SHUTDOWN/REFUELING MATRIX JAF 2-27 COLD SHUTDOWN/REFUELING MATRIX EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • The pre-calculated effluent monitor values presented in EAL #1 (Table R 1) should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes.

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
a. > 1000 mRem TEDE OR
b. > 5000 mRem CDE Thyroid OR
3. Field survey results at or beyond the site boundary indicate EITHER:
a. Gamma (closed window) dose rates> 1000 mR/hr are expected to continue for~ 60 minutes.

OR

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT RG1 (cont)

Emergency Action Level (EAL) (cont):

Table R1 Effluent Monitor Thresholds Effluent Monitor General Emergency Stack 7880 mR/hr Turb Bldg Exh 2.44 mR/hr Radw Bldg Exh 4.74 mR/hr 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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. JAFNPP Technical Specifications Section 4.1.1, Figure 4.4-1
3. OP-31 Process Radiation Monitoring Systems
4. DVP-01.02 Offsite Dose Calculation Manual
5. NEI 99-01 Rev 6, AG1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • The pre-calculated effluent monitor values presented in EAL #1 (Table R1) should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes.

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
a. > 100 mRem TEDE OR
b. > 500 mRem CDE Thyroid OR
3. Field survey results at or beyond the site boundary indicate EITHER:
a. Gamma (closed window) dose rates> 100 mR/hr are expected to continue for~ 60 minutes.

OR

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

July 2019 JAF 2-30 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS1 (cont)

Emergency Action Level (EAL) (cont):

Table R1 Effluent Monitor Thresholds Effluent Monitor Site Area Emergency Stack 788 mR/hr Turb Bldg Exh 0.244 mR/hr Radw Bldg Exh 0.474 mR/hr 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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RG1.

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. JAFNPP Technical Specifications Section 4.1.1, Figure 4.4-1
3. OP-31 Process Radiation Monitoring Systems
4. DVP-01.02 Offsite Dose Calculation Manual
5. NEI 99-01 Rev 6, AS1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA1 Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mRem TEDE or 50 mRem thyroid COE.

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • The pre-calculated effluent monitor values presented in EAL #1 (Table R 1) should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
1. Readings on ANY Table R1 Effluent Monitor> Table R1 value for~ 15 minutes.

OR

2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
a. > 10 mRem TEDE OR
b. > 50 mRem CDE Thyroid OR
3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
a. 10 mRem TEDE for 60 minutes of exposure OR
b. 50 mRem CDE Thyroid for 60 minutes of exposure

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 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.

OR

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

Table R1

  • Effluent Monitor Thresholds Effluent Monitor Alert Stack 78.8 mR/hr Turb Bldg Exh N/A
  • Radw Bldg Exh Basis:

N/A 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). r 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 TEOE dose is set at 1% of the EPA PAG of 1000 mRem while the 50 mRem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.

Escalation of the emergency classification level would be via IC RS1 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA1 (cont)

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. JAFNPP Technical Specifications Section 4.1.1, Figure 4.4-1
3. OP-31 Process Radiation Monitoring Systems
4. DVP-01.02 Offsite Dose Calculation Manual
5. NEI 99-01 Rev 6, AA1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
  • Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes .
  • 1. Reading on the Liquid Radwaste Effluent Monitor (1 ?RM-350) > 2 times hi-hi trip for~ 60 minutes.

OR

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

Table R1 Effluent Monitor Thresholds Effluent Monitor Unusual Event 1--------

Stack 0.451 mR/hr (High Range Monitor)

Rx Bldg Exh 9.50E+05 cpm (Low Range Monitor)

Turb Bldg Exh 6. 72E+05 cpm (Low Range Monitor)

Refuel Floor Exh 9.28E+05 cpm (Low Range Monitor)

OR

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU1 (cont)

Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

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

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

The effluent monitor listed is normally used for planned discharges.

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 RA 1.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU1 (cont)

Basis Reference(s):

1. EP-EAL-0637, Calculation of FitzPatrick Table R-1 EAL Threshold Values
2. DVP-01.02 Offsite Dose Calculation Manual
3. OP-31 Process Radiation Monitoring Systems
4. NEI 99-01 Rev 6, AU1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RG2 Initiating Condition:

Spent fuel pool level cannot be restored to at least 1.00 foot for 60 minutes or longer.

Operating Mode Applicability:

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

Note:

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

Spent fuel pool level cannot be restored to at least 1.00 foot as indicated on 19Ll-60A or 19Ll-60B 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. EOP-5 Secondary Containment Control
2. FSG-005, Alternate Spent Fuel Pool Makeup and Cooling
2. NEI 99-01 Rev 6, AG2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RS2 Initiating Condition:

Spent fuel pool level at 1.00 foot Operating Mode Applicability:

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

Lowering of spent fuel pool level to 1.00 foot as indicated on 19Ll-60A or 19Ll-60B.

Basis:

This IC addresses a significant loss ofspent 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.

2.

EOP-5 Secondary Containment Control FSG-005, Alternate Spent Fuel Pool Makeup and Cooling

3. NEI 99-01 Rev 6, AS2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT RA2 Initiating Condition:

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

Operating Mode Applicability:

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

1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

OR

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY Table R2 Radiation Monitor Alarm.

OR

3. Lowering of spent fuel pool level to 11.00 feet as indicated on 19Ll-60A or 19Ll-608 .

Table R2 Refuel Floor Radiation Monitors 1BRIA-051-12 Spent Fuel Pool (EPIC A-1229) 18RIA-051-14 New Fuel Vault (EPIC A-1231)

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 or a significant lowering of water level within the spent fuel pool.

July 2019 JAF 2-40 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT RA2 (cont)

Basis (cont):

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.

EAL #1 Basis This EAL escalates from RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

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

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

EAL #2 Basis This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

EAL #3 Basis:

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA2 (cont)

Basis Reference(s):

1. EOP-5 Secondary Containment Control
2. OP-32 Area Radiation Monitoring
3. JAFNPP EPG/SAG
4. FSG-005, Alternate Spent Fuel Pool Makeup and Cooling
5. NEI 99-01 Rev 6, AA2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU2 Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.

Operating Mode Applicability:

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

1. a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of*the following:
  • Inability to restore and maintain Spent Fuel Pool water level

> low water level alarm.

OR

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

AND

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

Table R2 Refuel Floor Radiation Monitors

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

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

This IC addresses a loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also July 2019 JAF 2-43 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU2 (cont)

Basis (cont):

indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

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

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

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

  • Escalation of the emergency classification level would be via IC RA2.

Basis Reference(s):

1. AOP-53 Loss of Spent Fuel Pool, Reactor Cavity or Equipment Storage Pit Water Level
2. OP-32 Area Radiation Monitoring
3. OP-30 Fuel Pool Cooling and Cleanup System
4. ARP 09-3-1-9 Fuel Pool Cool & Cln Up Trouble
5. AOP-68 Spent Fuel Pool Trouble
6. NEI 99-01 Rev 6, AU2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RAJ Initiating Condition:

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

Operating Mode Applicability:

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

Note:

  • If the equipment in the room or area listed in Table 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 in Table R3.

Table R3

  • OR Main Control Room - (by survey)

Central Alarm Station - (by survey)

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the areas in Table R4.

Areas with Ent Area

  • Reactor Building East Crescent
  • Reactor Building West Crescent
  • Reactor Building 272' Elevation Mode 3, 4, and 5
  • Reactor Building 300' Elevation
  • Relay Room
  • North Cable Room

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 (cont)

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal plant procedures. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR, therefore Table R4 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to

  • attain and maintain cold shutdown. This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

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

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

An emergency declaration is not warranted if any of the following conditions apply.

  • The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4 .

July 2019 JAF 2-46 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RA3 (cont)

Basis (cont):

  • The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
  • The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
  • The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

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

Basis Reference(s):

1. JAFNPP Safe Shutdown Analysis
2. NEI 99-01 Rev 6, AA3

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT RU3 Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability:

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

1. Offgas radiation ~ hi-hi alarm OR
2. Specific coolant activity> 2.0 µCi/gm 1-131 dose equivalent.

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.

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

  • Escalation of the emergency classification level would be via ICs FA 1 or the Recognition Category R ICs.

Basis Reference(s):

1. DVP-01.02 Offsite Dose Calculation Manual Specification 3.6.1
2. Technical Specification 3.7.5
3. Technical Specification 3.4.6
4. Technical Specification Bases 3.4.6
5. OP-31 Process Radiation Monitoring
6. NEI 99-01 Rev 6, SU3

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 Initiating Condition:

Loss of ANY two barriers AND Loss or Potential Loss of the third barrier.

Operating Mode Applicability:

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

Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.

Basis:

Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the General Emergency classification level each barrier is weighted equally.

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 Initiating Condition:

Loss or Potential Loss of ANY two barriers.

Operating Mode Applicability:

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

Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.

Basis:

Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the Site Area Emergency classification level, each barrier is weighted equally.

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 Initiating Condition:

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

Operating Mode Applicability:

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

Refer to Fission Product Barrier Loss and Potential Loss threshold values to determine barrier status.

Basis:

Fuel Cladding, RCS and Containment comprise the fission product barriers.

At the Alert classification level, Fuel Cladding and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Cladding or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Cladding or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.

  • Basis Reference(s):
1. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC1 Initiating Condition:

RCS Activity Operating Mode Applicability:

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

LOSS Coolant activity> 300 uCi/gm 1-131 dose equivalent.

Basis:

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

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a

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

There is no Potential Loss threshold associated with RCS Activity.

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS

1. SAOG entry required.

POTENTIAL LOSS

2. RPV water level cannot be restored and maintained> 0 inches (TAF).

OR

3. RPV water level cannot be determined.

Basis:

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

  • accordance with the EOP/SAOG program.

Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for SAOG entry. This is identified in the BWROG EPGs/SAGs when the phrase, "Enter SAOGs," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, "SAOG entry required," also accommodates the EOP need to enter SAOGs when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.

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

The RPV water level threshold is the same as RCS Barrier RC2 Loss threshold. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)

Basis (cont):

choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization

  • has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

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

  • considered reached as soon as it is apparent that the top of active fuel cannot be attained.

Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

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

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-2 RPV Control
3. EOP-7 RPV Flooding
4. EOP-3 Failure to Scram
5. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS Drywell radiation monitor reading> 1.8E+03 R/hr (1800 R/hr).

Basis:

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS

  • Barrier RC5 Loss Threshold since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two thresholds appropriately escalates the emergency classification level to a Site Area Emergency.

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

Basis Reference(s):

1. EP-EAL-0715, Criteria for Choosing Containment Radiation values Indicating: loss of fuel clad and potential loss of containment for Fitzpatrick Nuclear Power Station
2. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS

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

POTENTIAL LOSS

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

Basis:

Loss Threshold #1 Basis This threshold addresses any other factors that are to be used by the Emergency

  • Director in determining whether the Fuel Clad Barrier is lost.

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

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

LOSS

1. RPV water level cannot be restored and maintained> 0 inches (TAF)

OR

2. RPV water level cannot be determined.

Basis:

RPV values are actual levels, not indicated levels. Therefore, they may need level compensation depending on conditions. Compensated values may be used in accordance with the EOP/SAOG program.

This water level corresponds to the Top of Active Fuel (TAF) and is used in the EOPs to

  • indicate challenge to core cooling .

The RPV water level threshold is the same as Fuel Clad Barrier FC2 Potential Loss threshold. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC2 (cont)

Basis (cont):

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

Entry into the "Steam Cooling" leg of the EOP's would be an example of an inability to "restore and maintain" level above TAF resulting in this threshold being met.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level in order to reduce reactor power. Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA3 or MS3 will dictate the need for emergency classification.

  • There is no RCS Potential Loss threshold associated with RPV Water Level.

Basis Reference(s):

1.

2.

EP-1 EOP Entry and Use EOP-2 RPV Control

3. EOP-7 RPV Flooding
4. EOP-3 Failure to Scram
5. TSG-1 Parameter Assessment
6. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:

Primary Containment Pressure / Conditions Operating Mode Applicability:

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

LOSS

1. a. Primary Containment pressure> 2.7 psig.

AND

b. Primary Containment pressure rise is due to RCS leakage Basis:

The > 2.7 psig primary containment pressure is the Drywell high pressure setpoint which indicates a LOCA by automatically initiating ECCS.

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

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

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

There is no Potential Loss threshold associated with Primary Containment Pressure.

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-2 RPV Control
3. EOP-4 Primary Containment Control, Entry Conditions
4. FSAR Update Chapter 6 Emergency Core Cooling Systems
5. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 Initiating Condition:

RCS Leak Rate Operating Mode Applicability:

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

LOSS

1. UNISOLABLE Main Steam Line (MSL), HPCI, RWCU, RCIC, or Feedwater line break.

OR

2. Emergency RPV Depressurization is required.

POTENTIAL LOSS

3. UNISOLABLE primary system leakage that results in EITHER of the following:
a. Secondary Containment area temperature > EOP-5 Maximum Normal Operating Limit.
  • Basis:

OR

b. Secondary Containment area radiation > EOP-5 Maximum Normal Operating Limit.

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Classification of a system break over system leakage is based on information available to the Control Room from the event. Indications that should be considered are:

  • Reports describing magnitude of steam or water release.
  • Use of system high flow alarms/ indications, if available,
  • Significant changes in makeup requirements,
  • Abnormal reactor water level changes in response to the event.

The use of the above indications provides the Control Room the bases to determine that the on going event is more significant than the indications that would be expected from system leakage and therefore should be considered a system break .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis (cont):

Loss Threshold #1 Basis Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, the RCS barrier Loss threshold is met.

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.

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

Potential Loss Threshold #3 Basis Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Radiation values in areas that contain piping associated with main steam line, RCIC, HPCI, Feedwater, RWCU, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is considered to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis (cont):

Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier CT6 Loss Threshold

  1. 1 (following automatic or manual isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Basis Reference(s):

1. FM-29A Main Steam System Flow Diagram
2. FM-29B Main Steam System Flow Diagram
3. FM-25A High Pressure Coolant Injection System Flow Diagram
4. FM-22A Reactor Core Isolation Cooling System Flow Diagram
5. FM-34A Feedwater System Flow Diagram
6. EP-1 EOP Entry and Use
7. EOP-2 RPV Control
8. EOP-3 Failure to Scram
9. EOP-4 Primary Containment Control
10. EOP-5 Secondary Containment Control
11. EOP-6 Radioactivity Release Control
12. EOP-7 RPV Flooding 13 . NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RCS Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS Drywell radiation monitor reading > 63 R/hr.

Basis:

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier FC5 Loss Threshold since it indicates a loss of the RCS Barrier only.

There is no RCS Potential Loss threshold associated with Primary Containment Radiation .

  • Basis Reference(s):

1.

2.

EP-EAL-0515, Criteria for Choosing Drywell Radiation Monitor Reading Indicative of Loss of the RCS Barrier for Fitzpatrick Station NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS

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

POTENTIAL LOSS

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

Basis:

Loss Threshold #1 Basis

-This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS Barrier is lost.

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

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT2 Initiating Condition:

RPV Water Level Operating Mode Applicability:

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

POTENTIAL LOSS SAOG entry required.

Basis:

The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV Water Level. The Potential Loss requirement for entry into the Severe Accident Procedures (SAOGs) indicates adequate core cooling cannot be restored and maintained and that core damage is possible. Entry into SAOGs is in response to the inability to restore and maintain adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential

  • for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-2 RPV Control
3. EOP-3 Failure to Scram
4. EOP-7 RPV Flooding
5. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Primary Containment Pressure/ Conditions Operating Mode Applicability:

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

LOSS

1. UNPLANNED rapid drop in Primary Containment pressure following Primary Containment pressure rise.

OR

2. Primary Containment pressure response not consistent with LOCA conditions.

POTENTIAL LOSS

3. Primary Containment pressure > 56 psig.

OR

4. a. Primary Containment hydrogen concentration~ 6% .

AND

b. Primary Containment oxygen concentration ~ 5%.

OR

5. Heat Capacity Temperature Limit (HCTL) (EOP-11) exceeded.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Loss Threshold #1 and #2 Basis Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to Drywell spray or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. A pressure suppression bypass path would not be an indication of a

  • containment breach .

July 2019 JAF 2-66 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 (cont)

Basis (cont):

Potential Loss Threshold #3 Basis The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.

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

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

Basis Reference(s):

1. FSAR Update Section 5.2.3
2. EOP-4 Primary Containment Control
3. UFSAR 14.6.1.3.3
4. BWROG EPG/SAG Revision 3, Sections PC/G
5. FSAR section 5.2.3.14
6. FSAR Table 7.3-6
7. BWROG EPG/SAG Revision 3, Section 18
8. EOP-11 EOP and SAOG Graphs
9. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CTS Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

POTENTIAL LOSS Drywell radiation monitor reading> 1.8E+04 R/hr (18,000 R/hr).

Basis:

There is no Loss threshold associated with Primary Containment Radiation.

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

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20%

in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.

Basis Reference(s):

1. EP-EAL-0715, Criteria for Choosing Containment Radiation values Indicating: loss of fuel clad and potential loss of containment for Fitzpatrick Nuclear Power Station
2. NEI 99-01 Rev 6, Table 9-F-2
  • July 2019 JAF 2-68

/

EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 Initiating Condition:

Primary Containment Isolation Failure Operating Mode Applicability:

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

LOSS

1. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal.

OR

2. Intentional Primary Containment venting or purging per EOPs or SAOGs due to accident conditions.

OR

3. UNISOLABLE primary system leakage that results in EITHER of the following:
a. Secondary Containment area temperature > EOP-5 Maximum Safe Operating
  • Basis:

Limit.

OR

b. Secondary Containment area radiation > EOP-5 Maximum Safe Operating Limit.

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

Failure to isolate the leak, within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

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

Loss Threshold #1 Basis The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).

Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE Main Steam Line, or RCIC, HPIC, Feedwater line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths.

  • Examples of "downstream pathway to the environment" could be through the Turbine/Condenser, or direct release to the Turbine or Reactor Building.

July 2019 JAF 2-69 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system.

These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

Loss Threshold #2 Basis EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed .

Loss Threshold #3 Basis The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is considered to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in

  • conjunction with other indications (e.g. room flooding, high area temperatures, reports of July 2019 JAF 2-70 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected Main Turbine Control Valve closure) may indicate that a primary system is discharging into the Reactor Building.

In combination with RCS Barrier RC4 Potential Loss Threshold #3 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Basis Reference(s):

1. EOP-4 Primary Containment Control
2. EP-6 Post accident Containment Venting and Gas Control
3. EP-1 EOP Entry and Use
4. EOP-5 Secondary Containment Control
5. NEI 99-01 Rev 6, Table 9-F-2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT7 Initiating Condition:

Emergency Director Judgment.

Operating Mode Applicability:

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

LOSS

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

POTENTIAL LOSS

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

Basis:

Loss Threshold #1 Basis:

This threshold addresses any other factors that are to be used by the Emergency

  • Director in determining whether the Containment Barrier is lost.

Potential Loss Threshold #2 Basis:

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

Basis Reference(s):

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 Initiating Condition:

Prolonged loss of all offsite and all onsite AC power to emergency buses.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of ALL offsite and onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. EITHER of the following:
a. Restoration of at least one 4160 V emergency bus 10500 or 10600 in

< 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely .

  • Basis:

OR

b. RPV water level cannot be restored and maintained > -19 inches (MSCRWL).

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

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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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.

If mitigating strategies are effective in reestablishing emergency power to any of the buses listed, within the specified time, then declaration of this EAL is not warranted.

This EAL is not concerned with the source of the power as much as the loss of power to the listed buses.

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. Misc. Calculation JAF-CALC-89-012 "Determination of Required SBO Coping Duration Per NUMARC 8700"
2. OP-44 115 KV System
3. Drawing 71-002 AC Distribution
4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer
7. JAFNPP Plant-Specific Technical Guideline (PSTG)
8. EOP-2 RPV Control
9. EOP-7 RPV Flooding
10. AOP-49 Station Blackout
11. NEI 99-01 Rev 6, SG1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS 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 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 onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 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.

If mitigating strategies are effective in reestablishing emergency power to any of the buses listed, within the specified time, then declaration of this EAL is not warranted.

This EAL is not concerned with the source of the power as much as the loss of power to the listed buses.

Escalation of the emergency classification level would be via ICs RG1, FG1, MG1, or MG2 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 (cont)

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-45 345 KV System
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer
7. AOP-49 Station Blackout
8. NEI 99-01 Rev 6, SS1

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. AC power capability to 4160 V emergency buses 10500 and 10600 reduced to only one of the following power sources for~ 15 minutes.
  • Reserve Station Transformer T-2
  • Reserve Station Transformer T-3
  • EDGA
  • EOG B
  • EOG C
  • EOG D
  • Main Generator via T-4 AND
2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC MU1.

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 .

July 2019 JAF 2-77 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)

Basis (cont):

  • A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
  • Loss of offsite power and loss of all emergency power sources (e.g. onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

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. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-45 345 KV System
5. OP-22 Diesel Generator Emergency Power
  • 6.

7.

OP-45A Backfeeding Normal Station Service Transformer NEI 99-01 Rev 6, SA 1

James A. FitzPatrick Nuclear Power Plant Arinex Exelon Nuclear

  • RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU1 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

Loss of ALL offsite AC power capability to 4160 V emergency buses 10500 and 10600 for~ 15 minutes.

  • Reserve Station Transformer T-2
  • Reserve Station Transformer T-3
  • Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses.

This condition represents a potential reduction in the level of safety of the plant.

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. (e.g. unit cross-tie breakers)

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC MA 1.

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-45 345 KV System
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of ALL offsite and onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. Voltage is < 105 VDC on Vital DC buses 71 BCB-2A and 71 BCB-2B.

AND

3. ALL AC and Vital DC power sources in EALs #1 and #2 have been lost for
  • ?. 15 minutes.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when all EAL conditions are met.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
3. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
4. OP-22 Diesel Generator Emergency Power
5. OP-45A Backfeeding Normal Station Service Transformer
6. JAFNPP Plant-Specific Technical Guideline (PSTG)
7. EOP-2 RPV Control
8. EOP-7 RPV Flooding
9. AOP-49 Station Blackout
10. Drawing S71-068
11. OP-43A 125 voe System
12. ARP 09-8-1-20 125 voe Batt A Volt Lo
13. ARP 09-8-1-23 125 voe Batt B Volt Lo
  • 14.

15.

16.

AOP-45 Loss of DC Power System 'A' AOP-46 Loss of DC Power System 'B' NEI 99-01 Rev 6, SG8

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS2 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

Voltage is< 105 VDC on Vital DC buses 71 BCB-2A and 71 BCB-28 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. Drawing S71-068
2. OP-43A 125 voe System
3. ARP 09-8-1-20 125 voe Batt A Volt Lo
4. ARP 09-8-1-23 125 voe Batt B Volt Lo
5. AOP-45 Loss of DC Power System 'A'
6. AOP-46 Loss of DC Power System 'B'
7. NEI 99-01 Rev 6, SS8

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 Initiating Condition:

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

Operating Mode Applicability:

1, 2 Emergency Action Level (EAL):

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

~2.5%.

AND

2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power~ 2.5%.

AND

3. EITHER of the following conditions exist:
  • RPV water level cannot be restored and maintained > -19 inches (MSCRWL).

OR Basis:

Heat Capacity Temperature Limit (HCTL) (EOP-11) exceeded.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection, are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

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

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

Escalation of the emergency classification level would be via IC RG1 or FG1.

July 2019 JAF 2-83 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)

Basis Reference(s):

1. FSAR Update Section 7.2
2. EOP-3 Failure to Scram
3. EOP-2 RPV Control
4. EOP-4 Primary Containment Control
5. EOP-7 RPV Flooding
6. EOP-11 EOP and SAOG Graphs
7. NEI 99-01 Rev 6, SS5

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the Reactor Control Console 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. This action does not include manually driving in control rods or implementation of boron injection strategies.
1. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power~ 2.5%.

AND

2. Manual / ARI actions taken at the Reactor Control Console are not successful in shutting down the reactor as indicated by Reactor Power~ 2.5%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the Reactor Control Console to rapidly 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 scram, ARI). 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".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 (cont)

Basis (cont):

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

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

Basis Reference(s):

1. EP-3 Backup Control Rod Insertion
2. EOP-3 Failure to Scram
3. EOP-2 RPV Control

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 Initiating Condition:

Automatic or manual scram fails to shutdown the reactor.

Operating Mode Applicability:

1, 2 Emergency Action Level (EAL):

Note:

  • A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core. This action does not include manually driving in control rods or implementation of boron injection strategies.
1. a. Automatic scram did not shutdown the reactor as indicated by Reactor Power~ 2.5%.

AND

b. Subsequent manual / ARI action taken at the Reactor Control Console is successful in shutting down the reactor as indicated by Reactor Power < 2.5% .

OR

2. a. Manual scram did not shutdown the reactor as indicated by Reactor Power~ 2.5%.

AND

b. EITHER of the following:
1. Subsequent manual/ ARI action taken at the Reactor Control Console is successful in shutting down the reactor as indicated by Reactor Power < 2.5%.

OR

2. Subsequent automatic scram I ARI is successful in shutting down the reactor as indicated by Reactor Power< 2.5%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the Reactor Control Console or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)

Basis (cont):

EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the Reactor Control Console to shutdown the reactor (e.g., initiate a manual reactor scram/ARI). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

EAL#2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take other manual actions on the Reactor Control Console to shutdown the reactor (e.g., initiate a manual reactor scram / or initiating ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram, including ARI, is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the Reactor Control 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 scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the Reactor Control Console".

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the Reactor Control 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 FA 1, an Unusual Event declaration is appropriate for this event.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

  • If the signal generated as a result of plant work causes a plant transient that creates a real condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated .
  • July 2019 JAF 2-88 .EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)

Basis (cont):

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

Basis Reference(s):

1. EP-1 EOP Entry and Use
2. EOP-3 Failure to Scram
3. EOP-2 RPV Control
4. AOP-1 Reactor Scram
5. Technical Specifications section 3.3.1.1 RPS Instrumentation
6. NEI 99-01 Rev 6, SU5

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for~ 15 minutes.

Table M1 Control Room Parameters

2. ANY Table M2 transient in progress.

Table M2 Significant Transients

  • Auto/Manual runback > 25% thermal reactor power
  • Electric load rejection> 25% full electric load
  • Thermal Power oscillations > 10% (peak to peak)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 (cont)

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the 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, recorder source, or equivalent (e.g. camera) within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RS1 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA4 (cont)

Basis Reference(s):

1. FSAR Update Section 7 .16
2. FSAR Update Section 7.19
3. EOP-2 RPV Control
4. EOP-3 Failure to Scram
5. EOP-4 Primary Containment Control
6. EOP-5 Secondary Containment Control
7. EOP-6 Radioactivity Release Control
8. EOP-7 RPV Flooding
9. NEI 99-01 Rev 6, SA2

James A. FitzPatrick Nuclear Power Plarit Annex Exelon Nuclear RECOGNITION ~CATEGORY SYSTEM MALFUNCTIONS MU4 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

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

UNPLANNED event results in the inability to monitor ANY Table M1 parameter from within the Control Room for~ 15 minutes.

Table M1 Control Room Parameters

  • Reactor Power
  • RPV Water Level
  • Basis:

Primary Containment Pressure Torus Level Torus Temperature UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for any of the listed parameters cannot be determined from within the Control Room. This *situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital, recorder source, or equivalent (e.g. camera) within the Control Room .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4 (cont)

Basis (cont):

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50. 72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for any of the listed parameters are lost, then the ~bility to determine the values of 1other 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. FSAR Update Section 7.16
2. FSAR Update Section 7.19
3. NEI 99-01 Rev 6, SU2

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAS Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.

Operating Mode Applicability:

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

Note:

  • This EAL is only applicable to SAFETY SYSTEMs having two (2) or more trains.
  • If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred, then this emergency classification is not warranted.
  • If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
  • If a hazardous event occurs and it is determined that the conditions of MA5 are not met, then assess the event via HU3, HU4, or HU6.
1. a. The occurrence of ANY of the following hazardous events:

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

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.
  • AND
c. EITHER of the following:
  • Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

OR

  • Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAS (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.

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train 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 SAFETY SYSTEM train .

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS required for the current operaUng mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. In order to provide the appropriate context for consideration of an Alert classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has degraded performance for criteria 1.b of this EAL; commercial nuclear power plants are designed to be able to support single system issue*s without compromising public health and safety from radiological events.

Indications of degraded performance address damage to a SAFETY SYSTEM train that 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 .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MAS {cont}

Basis (cont):

Operators will make a determination of VISIBLE DAMAGE 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or RS1.

If a hazardous event occurs and 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

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 Initiating Condition:

RCS leakage for 15 minutes or longer.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for~ 15 minutes.

OR

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

OR

(

3. Leakage from the RCS to a location outside the Drywell > 25 gpm for~ 15 minutes.
  • Basis:

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

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. The assessment of this EAL may be based on the results of RCS leak rate calculation that may be necessary to ascertain whether the EAL has been exceeded. In this case, the 15-minute declaration period starts with the availability of the RCS leak rate calculation results that show the EAL to be exceeded (i.e., this is the time that the EAL information is first available).

These three EALs thus apply to leakage into the containment, a secondary-side system

  • or a location outside of containment.

July 2019 JAF 2-98 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)

Basis (cont):

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

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

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category R or F.

Basis Reference(s):

1. FSAR Update Section 4.10
2. NEI 99-01 Rev 6, SU4

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

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

1. Loss of ALL Table M3 onsite communication capabilities affecting the ability to perform routine operations.

OR

2. Loss of ALL Table M3 offsite communication capabilities affecting the ability to perform offsite notifications.

OR

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

Table M3 Communication Capabilities

  • System Page/Party System (Gaitronics)

Control Room/Portable Radio Plant Telephones (all VOiP, switched, non-switched)

Onsite X

X X

Offsite X

NRC X

Installed Out-of-Plant Cellular Phones X X X Plant Satellite Phones (Installed in CR and deployable) X X RECS X Dedicated Phone Lines (ENS) X X HPN and FTS 2001 X X Basis:

This IC addresses a significant loss of onsite, offsite, or NRC communication 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.) .

  • July 2019 JAF 2-100 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)

Basis (cont):

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

EAL #2 Basis Addresses a total loss of the communication methods used to notify all Offsite Response Organizations (OROs) of an emergency declaration. The Offsite Response Organizations (OROs) referred to here are listed in procedure EP-CE-114-100-F-05, JAF Notification Fact Sheet .

EAL #3 Basis Addresses a total loss of the communication methods used to notify the NRC of an emergency declaration.

Basis Reference(s):

1. NY State Emergency Operations Center
2. NY State Warning Point
3. Alternate State Warning Point
4. State Department of Health
5. SEMO Regional Office
6. Oswego County EOC
7. Oswego County E-911 Center (Warning Point)
8. Nine Mile Point Control Rooms
9. Nine Mile Point TSC and EOF
10. JAFNPP Control Room
11. JAFNPP TSC
12. JAFNPP EOF
13. SEMO Technical Resources 14 . NEI 99-01 Rev 6, SU6
  • July 2019 JAF 2-101 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 Initiating Condition:

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

Operating Mode Applicability:

4,5,D Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of ALL offsite and onsite AC power to 4160 V emergency buses 10500 and 10600.

AND

2. Failure to restore power to at least one 4160 V emergency bus 10500 or 10600 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.

If mitigating strategies are effective in reestablishing emergency power to any of the buses listed, within the specified time, then declaration of this EAL is not warranted.

This EAL is not concerned with the source of the power as much as the loss of power to the listed buses .

  • July 2019 JAF 2-102 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)

Basis (cont):

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.

Basis Reference(s):

1. Drawing 71-002 AC Distribution
2. OP-44 115 KV System
3. OP-45 345 KV System
4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
5. OP-22 Diesel Generator Emergency Power
6. OP-45A Backfeeding Normal Station Service Transformer
7. NEI 99-01 Rev 6, CA2
  • July 2019 JAF 2-103 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU1 Initiating Condition:

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

Operating Mode Applicability:

4,5, 0 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 4160 V emergency buses 10500 and 10600 reduced to only one of the following power sources for~ 15 minutes.
  • Reserve Station Transformer T-2
  • Reserve Station Transformer T-3
  • EOGA
  • EOG B
  • EOG C
  • EOG 0 AND
2. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

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

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

July 2019 JAF 2-104 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU1 (cont)

Basis (cont):

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

  • A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
  • Loss of offsite power and loss of all emergency power sources (e.g. onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

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 CA 1.

Basis Reference(s):

1. OP-44 115 KV System
2. Drawing 71-002 AC Distribution
  • 3.

4.

5.

6.

OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution OP-45 345 KV System OP-22 Diesel Generator Emergency Power OP-45A Backfeeding Normal Station Service Transformer

7. NEI 99-01 Rev 6 CU2
  • July 2019 JAF 2-105 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:

Hazardous event affecting SAFETY SYSTEM required for the current operating mode.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • This EAL is only applicable to SAFETY SYSTEMs having two (2) or more trains.
  • If the affected SAFETY SYSTEM train was already inoperable before the hazardous event occurred, then this emergency classification is not warranted.
  • If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
  • If a hazardous event occurs and it is determined that the conditions of CA2 are not met, then assess the event via HU3, HU4, or HU6.
1. a. The occurrence of ANY of the following hazardous events:

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

b. Event damage has caused indications of degraded performance to one train of a SAFETY SYSTEM required by Technical Specifications for the current operating mode.

AND

c. EITHER of the following:
  • Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode.

OR

  • Event damage has resulted in VISIBLE DAMAGE to a second train of the SAFETY SYSTEM required by Technical Specifications for the current operating mode .
  • July 2019 JAF 2-106 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (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.

  • EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train 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 SAFETY SYSTEM train .

  • This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS required for the- current operating mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. In order to provide the appropriate context for consideration of an Alert classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has degraded performance for criteria 1.b of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance address damage to a SAFETY SYSTEM train that 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 .

  • July 2019 JAF 2-107 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (cont)

Basis (cont):

Operators will make a determination of VISIBLE DAMAGE 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or RS1.

If a hazardous event occurs and 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 CA6
  • July 2019 JAF 2-108 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 Initiating Condition:

Loss of Vital DC power for 15 minutes or longer.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

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

Voltage is< 105 VDC on required Vital DC buses 71 BCB-2A and 71 BCB-2B 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 arid 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 Bis 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.

  • July 2019 JAF 2-109 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU3 (cont)

Basis Reference(s):

1. Drawing S71-068
2. OP-43A 125 voe System
3. ARP 09-8-1-20 125 voe Batt A Volt Lo
4. ARP 09-8-1-23 125 voe Batt AB Volt Lo
5. AOP-45 Loss of DC Power System 'A'
6. AOP-46 Loss of DC Power System 'B'
7. NEI 99-01 Rev 6, CU4

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:

Loss of all onsite or offsite communication capabilities.

Operating Mode Applicability:

4,5, D Emergency Action Level (EAL):

1. Loss of ALL Table C1 onsite communication capabilities affecting the ability to perform routine operations.

OR

2. Loss of ALL Table C1 offsite communication capabilities affecting the ability to perform offsite notifications.

OR

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

Table C1 Communication Capabilities

  • System Page/Party System (Gaitronics)

Control Room/Portable Radio Plant Telephones (all VOiP, switched, non-switched)

Onsite X

X X

Offsite X

NRC X

Installed Out-of-Plant Cellular Phones X X X Plant Satellite Phones (Installed in CR and deployable) X X RECS X Dedicated Phone Lines (ENS) X X HPN and FTS 2001 X X Basis:

This IC addresses a significant loss of onsite, offsite, or NRC communication 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.) .

July 2019 JAF2-111 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 {cont)

Basis (cont):

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

EAL #2 Basis Addresses a total loss of the communication methods used to notify all Offsite Response Organizations (OROs) of an emergency declaration. The Offsite Response Organizations (OR Os) referred to here are listed in procedure EP-CE-114-1 OO-F-05, JAF Notification Fact Sheet.

EAL #3 Basis Addresses a total loss of the communication methods used to notify the NRC of an emergency declaration.

Basis Reference(s):

1. NY State Emergency Operations Center
2. NY State Warning Point
3. Alternate State Warning Point
4. State Department of Health
5. SEMO Regional Office
6. Oswego County EOC
7. Oswego County E-911 Center (Warning Point)
8. Nine Mile Point Control Rooms
9. Nine Mile Point TSC and EOF
10. JAFNPP Control Room
11. JAFNPP TSC
12. JAFNPP EOF
13. SEMO Technical Resources
14. NEI 99-01 Rev 6, CU5
  • July 2019 JAF 2-112 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 Initiating Condition:

Inability to maintain the plant in cold shutdown.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.
1. UNPLANNED rise in RCS temperature> 212 °F for> Table C2 duration.

Table C2 RCS Heat-up Duration Thresholds

  • RCS Status Intact Not Intact.

Containment Closure Status Not Applicable Established Heat-up Duration 60 minutes*

20 minutes*

Not Established 0 minutes

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

OR

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

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions .

July 2019 JAF 2-113' EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGO-RY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA5 (cont)

Basis (cont):

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature rise.

The RCS Heat-up Duration Thresholds table also addresses a rise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the

  • temperature rise without a substantial degradation in plant safety.

Finally, in the case where there is a rise in RCS temperature, the RCS is not intact , and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL #2 provides a pressure-based indication of RCS heat-up.

Escalation of the emergency classification level would be via IC CS6 or RS1.

Basis Reference(s):

1. Technical Specifications Table 1.1-1
2. AOP-30 Loss of Shutdown Cooling
3. OP-13D RHR-Shutdown Cooling
4. Technical Specifications Section 3.61.1 and 3.6.4.1
5. NEI 99-01 Rev 6, CA3
  • July 2019 JAF 2-114 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 Initiating Condition:

UNPLANNED rise in RCS temperature Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when heat removal function is available does not warrant classification.
1. UNPLANNED rise in RCS temperature> 212 °F.

OR

2. Loss of the following for~ 15 minutes.
  • Basis:

ALL RCS temperature indications AND ALL RPV water level indications UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA5.

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.) .

  • July 2019 JAF 2-115 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU5 (cont)

Basis (cont):

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

EAL #1 Basis This 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 Basis This reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA6 based on an inventory loss or IC CA5 based on exceeding plant configuration-specific time criteria.

Basis Reference(s):

1. Technical Specifications Table 1.1-1
2. AOP-30 Loss of Shutdown Cooling
3. Drawing S02-069
4. NEI 99-01 Rev 6, CU3
  • July 2019 JAF 2-116 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Initiating Condition:

Loss of RPV inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. a. RPV water level < 0 inches (TAF) for~ 30 minutes.

AND

b. ANY Table C4 Containment Challenge Indication.

OR

2. a. RPV water level cannot be monitored for~ 30 minutes.

AND

b. Core uncovery is indicated by ANY of the following:
  • Table C3 indication of a sufficient magnitude to indicate core uncovery.

OR

~ 3 R/hr.

AND

c. ANY Table C4 Containment Challenge Indication .
  • July 2019 JAF 2-117 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Emergency Action Level (EAL) (cont):

Table C3 Indications of RCS Leakage

  • UNPLANNED Drywell equipment drain sump level rise*
  • UNPLANNED Drywell floor drain sump level rise*
  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory.

Table C4 Containment Challenge Indications Primary Containment Hydrogen Concentration UNPLANNED rise in primary containment pressure CONTAINMENT CLOSURE not established*

~ 6% and Oxygen ~

Secondary Containment area radiation > ANY Maximum Safe Operating Limit 5%

(EOP-5)

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

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

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

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions .

  • July 2019 JAF 2-118 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Basis (cont):

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guidelines (PAG) exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

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

  • In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access.

During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

EAL #2 Basis The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV .

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Basis (cont):

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.

Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G
2. EOP-4a Primary Containment Gas Control
3. FSAR section 5.2.3.14
4. FSAR Update Table 5.2-1
5. Technical Support Guideline-1 (TSG-1) Parameter A~sessment.
6. FSAR Update Section 4.10.3
7. OP-13D RHR-Shutdown Cooling
8. EOP-5 Secondary Containment Control
  • 9.

10.

11.

12.

Technical Specifications Sections 3.6.1.1 and 3.6.4.1 EOP-2 RPV Control EP-EAL-0506 Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Fitzpatrick Station NEI 99-01 Rev 6, CG1

  • July 2019 JAF 2-120 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 Initiating Condition:

Loss of RPV inventory affecting core decay heat removal capability.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. With CONTAINMENT CLOSURE not established, RPV water level

< 120.5 inches.

OR

2. With CONTAINMENT CLOSURE established, RPV water level< 0 inches (TAF).

OR

3. a. RPV water level cannot be monitored for~ 30 minutes AND
b. Core uncovery is indicated by ANY of the following:
  • Table C3 indication of a sufficient magnitude to indicate core uncovery.

OR

  • July 2019 JAF 2-121 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)

Emergency Action Level (EAL) (cont):

Table C3 Indications of RCS Leakage

  • UNPLANNED Drywell equipment drain sump level rise*
  • UNPLANNED Drywell floor drain sump level rise*
  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment (primary or secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

EAL #1 and #2 Basis The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs #1 and #2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

  • July 2019 JAF 2-122 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (co_nt)

Basis (cont):

EAL #3 Basis The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG6 or RG1 .

  • July 2019 JAF 2-123 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)

Basis Reference(s):

1. Technical Support Guideline-1 (TSG-1) Parameter Assessment
2. FSAR Update Section 4.10.3
3. OP-13D RHR-Shutdown Cooling
4. EOP-5 Secondary Containment Control
5. Technical Specifications Sections 3.6.1.1 and 3.6.4.1
6. EOP-2 RPV Control
7. EP-EAL-0506 Estimation Of Radiation Monitor Readings Indicating Core Uncovery During Refueling Fitzpatrick Station
8. NEI 99-01 Rev 6, CS1
  • July 2019 JAF 2-124 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 Initiating Condition:

Loss of RPV inventory.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. Loss of RPV inventory as indicated by level< 126.5 inches.

OR

2. a. RPV water level cannot be monitored for~ 15 minutes.

AND

b. Loss of RPV inventory per Table C3 indications .

Table C3 Indications of RCS Leakage UNPLANNED Drywell equipment drain sump level rise*

UNPLANNED Drywell floor drain sump level rise*

  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory.

Basis:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses 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 .

  • July 2019 JAF 2-125 EP-AA-1014 Addend.um 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 (cont)

Basis (cont):

EAL#1 Basis A lowering of water level below 126.5 inches indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery.

Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA5.

EAL #2 Basis The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS6 If the RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS6.

Basis Reference(s):

1. Technical Specifications Table 3.3.5.1.-1
2. Drawing S02-069
3. FSAR Update Section 4.10.3
4. OP-13D RHR-Shutdown Cooling
5. NEI 99-01 Rev 6, CA1
  • July 2019 JAF 2-126 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 Initiating Condition:

UNPLANNED loss of RPV inventory for 15 minutes or longer.

Operating Mode Applicability:

4,5 Emergency Action Level (EAL):

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. UNPLANNED loss of reactor coolant results in the inability to restore and maintain RPV water level above the procedurally established lower limit for?. 15 minutes.

OR

2. a. RPV water level cannot be monitored.

AND

b. Loss of RPV inventory per Table C3 indications.

Table C3 Indications of RCS Leakage

  • UNPLANNED Drywell equipment drain sump level rise*
  • UNPLANNED Drywell floor drain sump level rise*
  • UNPLANNED Reactor Building equipment sump level rise*
  • UNPLANNED Reactor Building floor drain sump level rise*
  • UNPLANNED Torus level rise*
  • UNPLANNED RPV make up rate rise*
  • Observation of leakage or inventory loss
  • Rise in level is attributed to a loss of RPV inventory .
  • July 2019 JAF 2-127 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)

Basis:

UNPLANNED:* A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV water level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

The procedurally established lower limit is not an operational band established above the procedural limit to allow for operator action prior to exceeding the procedural limit, but it is the procedurally established lower limit.

Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering 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 This recognizes that the minimum required RPV water level can change several times

  • during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL #2 Basis This addresses a condition where all means to determine RPV water level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be- evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA6 or CA5 .

  • July 2019 JAF 2-128 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)

Basis Reference(s):

1. Drawing 802-069
2. FSAR Update Section 4.10.3
3. OP-130 RHR-Shutdown Cooling
4. Technical Support Guideline-1 (TSG-1) Parameter Assessment
5. OP-658 Shutdown Operation
6. EOP-2 RPV Control
7. NEI 99-01, Rev. 6 CU1
  • July 2019 JAF 2-129 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:

HOSTILE ACTION within the PROTECTED AREA.

Operating Mode Applicability:

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

A notification from the Security Supervisor that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area) .

  • HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

July 2019 JAF 2-130 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 (cont)

Basis (cont):

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (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 and HG7.

Basis Reference(s):

1. JAFNPP Safeguards Contingency Plan
  • 2.

3.

AOP-70 Security Threat NEI 99-01 Rev 6, HS1

  • July 2019 JAF 2-131 EP-AA-1014 Addendum 3 (Rev. 2)

1 James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 Initiating Condition:

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

Operating Mode Applicability:

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

1. A validated notification from NRC of an aircraft attack threat< 30 minutes from the site.

OR

2. Notification by the Security Supervisor 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

  • July 2019 JAF 2-132 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)

Basis (cont):

PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE.

Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72 .

EAL #1 Basis Addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with AOP-70, Security Threat.

EAL #2 Basis Applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Escalation of the emergency classification level would be via IC HS1 .

July 2019 JAF 2-133 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)

Basis Reference(s):

1. JAFNPP Safeguards Contingency Plan
2. AOP-70 Security Threat
3. NEI 99-01 Rev 6, HA1
  • July 2019

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:

Confirmed SECURITY CONDITION or threat.

Operating Mode Applicability:

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

1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.

OR

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

OR

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

Basis:

SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site

    • personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety.

Security events which do not meet one of these EALs are adequately addressed by the

  • July 2019 JAF 2-135 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 (cont)

Basis (cont):

requirements of 10 CFR § 73. 71 or 10 CFR § 50. 72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA 1, and HS1.

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 Offsite Response Organizations (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 AOP-70, Security 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. JAFNPP Safeguards Contingency Plan
2. AOP-70 Security Threat
3. SY-AA-101-132, Security Assessment and Response to Unusual Activities
4. NEI 99-01 Rev 6, HU1
  • July 2019 JAF 2-136 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY-HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition:

Inability to control a key safety function from outside the Control Room.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
1. A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the Control Room.

AND

2. Control of ANY Table H1 key safety function is not reestablished in< 30 minutes.
  • Table H1 Safety Functions
  • Reactivity Control (ability to shut down the reactor and keep it shutdown)
  • RPV Water Level (ability to cool the core)
  • RCS Heat Removal (ability to maintain heat sink)

Basis:

The time period to establish control of the plant starts when either:

a. Control of needed safety functions 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 barrier within a relatively short period of time .

  • July 2019 JAF 2-137 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 (cont)

Basis (cont):

The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 30 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Escalation of the emergency classification level would be via IC FG1 or CG6.

Basis Reference(s):

1. AOP-43 Plant Shutdown from Outside the Control Room
2. NEI 99-01, Rev 6 HS6
  • July 2019 JAF 2-138 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2 Initiating Condition:

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

Operating Mode Applicability:

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

A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per AOP-43, Plant Shutdown from Outside the Control Room.

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. AOP-43 Plant Shutdown from Outside the Control Room
2. NEI 99-01, Rev 6 HA6
  • July 2019 JAF 2-139 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition:

FIRE potentially degrading the level of safety of the plant.

Operating Mode Applicability:

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

Note:

  • The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
  • Escalation of the emergency classification level would be via IC CA2 or MAS.
1. A FIRE in ANY Table H2 area is not extinguished in< 15 minutes of ANY of the following FIRE detection indications:
  • Report from the field (i.e., visual observation)
  • Receipt of multiple (more than 1) fire alarms or indications
  • Field verification of a single fire alarm Table H2 Areas
  • Reactor Building (when inerted the Drywell is exempt)
  • Control Room I Relay Room / Cable Run Rooms/ Cable Spreading Room
  • Electric Bays
  • Control Room AC Equipment Room
  • Control Room Chiller Room
  • Battery Rooms I Battery Room Corridor only
  • Cable Tunnels

OR

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

AND

b. The existence of a FIRE is not verified in< 30 minutes of alarm receipt.

OR

  • July 2019 JAF 2-140 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Emergency Action Level (EAL) (cont):

3. A FIRE within the plant PROTECTED AREA not extinguished in< 60 minutes of the initial report, alarm or indication.

OR

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

Basis:

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) : A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

  • This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

EAL #1 Basis The intent of the 15 minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarms, indication or report.

A fire in a Battery Charging Room should be evaluated based on its effects on the 125 VDC Vital Bus distribution system against EALs MS2 and CU3.

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 .

  • July 2019 JAF 2-141 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Basis (cont):

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30 minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30 minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

EAL #3 Basis In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60 minutes may also potentially degrade the level of plant safety.

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 July 2019 JAF 2-142 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Basis (cont):

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. FSAR Update Section 12.3
2. JAFNPP Safe Shutdown Analysis
3. NEI 99-01, Rev 6 HU4
  • July 2019 JAF 2-143 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition:

Seismic event greater than OBE levels.

Operating Mode Applicability:

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

Note:

  • For emergency classification if EAL # 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Emergency Director in ~ 15 minutes of the event.

OR

2. When Seismic Monitoring Equipment is not available:
  • a. Control Room personnel feel an actual or potential seismic event.

AND

b. ANY one of the following confirmed in ~ 15 minutes of the event:
  • The earthquake resulted in Modified Mercalli Intensity (MMI) 2:. VI and occurred

~ 3.5 miles of the plant.

  • The earthquake was magnitude 2:_ 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 a manner deemed appropriate by the Shift Manager or Emergency Director .
  • July 2019 JAF 2-144 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)

Basis:

EAL#1 Basis This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (QBE) 1 . An earthquake greater than an QBE but less than a Safe Shutdown Earthquake (SSE) 2 should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

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. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

EAL #2 Basis 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 guidance recognizes that it may cause the site to declare an Unusual Event while another site, similarly affected but with readily available QBE 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.

1 An QBE 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.

July 2019 JAF 2-145 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)

Basis Reference(s):

1. FSAR Update Section 2.6 Engineering Seismology
2. AOP-14 Earthquake
3. US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions
4. NEI 99-01, Rev 6 HU2
  • July 2019 JAF 2-146 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAS Initiating Condition:

Gaseous r~lease 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 a Table H3 area.

Areas with Ent Area Reactor Building East Crescent Reactor Building West Crescent Reactor Building 272' Elevation Reactor Building 300' Elevation Relay Room Mode 3, 4, and 5

  • North Cable Room 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.

  • July 2019 JAF 2-147 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAS (cont)

Basis (cont):

Assuming all plant equipment is operating as designed, normal operation is capable a

from the Main Control Room (MCR). The plant is also able to transition into hot shutdown condition from the MCR, therefore Table H3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.

This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

This Table does not include the Control Room since adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect and the gaseous release preclude the ability to place shutdown cooling in service. The emergency classification is not contingent upon whe_ther entry is actually necessary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

An emergency declaration is not warranted if any of the following conditions apply.

  • The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
  • The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
  • The action for which room/area entry is required is of an administrative or record
  • keeping nature (e.g., normal rounds or routine inspections) .

July 2019 JAF 2-148 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAS (cont)

Basis (cont):

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that generate smoke, that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

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

Basis Reference(s):

1. JAFNPP Safe Shutdown Analysis
2. NEI 99-01, Rev 6 HAS
  • July 2019 JAF 2-149 EP-AA-1014 Addendum 3 (Rev. 2)

_J

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 Initiating Condition:

Hazardous Event Operating Mode Applicability:

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

Note:

  • EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
  • Escalation of the emergency classification level would be via IC CA2 or MA5.
1. Tornado strike within the PROTECTED AREA.

OR

2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode .
3. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).

OR

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

OR

5. Intake Water Level> 255 feet.

OR

6. ESW intake bay water level ,:5. 237 feet.

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.

July 2019 JAF 2-150 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)

Basis (cont):

EAL #1 Basis Addresses a tornado striking (touching down) within the Protected Area.

EAL#2 Basis Addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

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

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 The high lake level is based upon the revised design flood level for the screenwell interior walls and gates.

EAL #6 Basis The low level of~ 237 feet is selected since IAW AOP-56, Intake Water Level Trouble at 237.5 feet there is adequate time to reduce power (scraming the reactor), remove the second CW pump and trend intake water level as mitigating actions. The ESW pumps are declared inoperable (Alert threshold) at 236.5 feet, so 237 feet allows for mitigating action to be taken prior to declaration and is above the Alert threshold allowing for escalation between the Unusual Event and the Alert thresholds.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, Hor C .

  • July 2019 JAF 2-151 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)

Basis Reference{s):

1. FSAR Section 2.4.3
2. Safety Evaluation JAF-SE-93-034 "Evaluation of Maximum and Minimum Water Levels at Screenwell for Safe Operation of Class I Equipment"
3. NEI 99-01, Rev 6 HU3
4. AOP-56, Intake Water Level Trouble
  • July 2019 JAF 2-152 EP-AA-1014 Addendum 3 (Rev. 2)

J

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG7 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY.

Operating Mode Applicability:

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

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

Basis:

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

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the. owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.

Basis Reference(s):

1. NEI 99-01, Rev 6 HG7
  • July 2019 JAF 2-153 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS7 Initiating Condition:

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

Operating Mode Applicability:

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

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to

  • achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage agajnst 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
  • July 2019 JAF 2-154 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA7 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of an ALERT.

Operating Mode Applicability:

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

Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts

  • that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert.

Basis Reference(s):

1. NEI 99-01, Rev 6 HA?
  • July 2019 JAF 2-155 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU7 Initiating Condition:

Other conditions exist which in the judgment of the Emergency Director warrant declaration of an UNUSUAL EVENT.

Operating Mode Applicability:

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

Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an UNUSUAL EVENT.

  • Basis Reference(s):
1. NEI 99-01, Rev 6 HU7
  • July 2019 JAF 2-156 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability:

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

Damage to*a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading> ANY Table E-1 values:

Table E-1 Radiation Reading Overpack Serial Overpack Average Overpack Overpack Number Surface Dose Serial Average Surface Rates mrem/hr Number Dose Rates (gamma+neutron) mrem/hr HI-STORM (gamma+neutron) 100S (XXX)

HI-STORM 80 on the side 100S S/N -0307, 600 on the side 20 on the top S/N - 15, 16, 17 0308, 0309, 32 at the inlet and 60 on the top 0310, 0311, outlet vent 0312, 0679, ducts 0680, 0681, HI-STORM 100 on the side 0682, 0683, 1oos (232) 0690, 0691, 20 on th*e top 0692, 0693, S/N - 0169, 0694, 0695, 90 at the inlet and 0170, 0171 outlet vent 1172, 1173, ducts 1174, 1175, 1176, 1177 HI-STORM 220 on the side 100S (XXX) 40 on the top S/N - 0186, 0187, 0188 Basis:

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

  • July 2019 JAF 2-157 EP-AA-1014 Addendum 3 (Rev. 2)

James A. FitzPatrick Nuclear Power Plant Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 (cont)

Basis (cont):

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The word cask, as used in this EAL, refers to the storage container in use at the site for dry storage of irradiated fuel. The issues of concern are the creation of a potential or actual release path to the environment, degradation of any fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate

  • limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSls are covered under ICs HU1 and HA1.

Basis Reference(s):

1. NEI 99-01, Rev 6 E-HU1
2. ISFSI Certificate Of Compliance Amendment No.'s 0, 1, 2, 5, and 8 .
  • July 2019 JAF 2-158 EP-AA-1014 Addendum 3 (Rev. 2)

ATTACHMENT 4 Radiological Emergency Plan Document Revision EP-AA-1014, Addendum 3, Appendix 1, Revision O, "JAF EAL Wallboard" J

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ATTACHMENT 5 Radiological Emergency Plan Document Revision EP-AA-1014, Addendum 3, Appendix 1, Revision 1, "JAF EAL Wallboard"

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