Text

Radiological Emergency Plan Addendum Revision
	ML17319A419
Person / Time
Site:	LaSalle
Issue date:	11/15/2017
From:	Gullott D; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	RS-17-162
	Download: ML17319A419 (169)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office 10 CFR 50.54(q)(5) 10 CFR 50.4 10 CFR 72.44(f)

RS-17-162 November 15, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NFP-11 and NFP-18 NRC Docket Nos. 50-373, 50-374, and 72-70

Subject:

Exelon Nuclear Radiological Emergency Plan Addendum Revision In accordance with 10 CFR 50.4(b)(5), "Emergency Plan and related submissions," Exelon Generation Company, LLC (Exelon) is submitting an Emergency Plan Addendum revision for LaSalle County Station (LaSalle) as listed in the table below.

Document Revision Title EP-AA-1005, Addendum 3 3 Emergency Action Levels for LaSalle Station The changes to the Emergency Plan Addendum were evaluated under the requirements of 10 CFR 50.54(q) and were determined not to result in a reduction in the effectiveness of the Emergency Plan for LaSalle. This notification is being submitted within 30 days of implementation of the changes as required by 10 CFR 50.54(q)(5). The changes continue to meet the applicable planning standards established in 10 CFR 50.47(b) and 10 CFR 50, Appendix E.

In addition, as required by 10 CFR 50.54(q)(5), this submittal includes a summary analysis of the changes to the Emergency Plan Addendum (Attachment 1).

This submittal also satisfies the reporting requirements associated with 10 CFR 72.44(f),

which stipulates that within six months after any change is made to the Emergency Plan, the licensee shall submit a report containing a description of the changes to the Director, Division of Spent Fuel Management.

U.S. Nuclear Regulatory Commission Emergency Plan Addendum Revision Docker Nos. 50-373, 50-374, and 72-70 November 15, 2017 Page 2 A copy of the revised Emergency Plan Addendum and supporting change summary analysis are included in the attachments to this letter.

There are no regulatory commitments in this submittal.

If you have any questions or require additional information, please contact Amy Hambly at (630) 657-2808.

Respectfully, David M. Gullatt Manager, Licensing Exelon Generation Company, LLC Attachments: 1) 1o CFR 50.54(q)(5} Procedure Change Summary Analysis

2) EP-AA-1005, Addendum 3, Revision 3, "Emergency Action Levels for LaSalle Station" cc: w/ Attachment 1 only Regional Administrator - NRC Region Ill Director, NRC Division of Spent Fuel Management, ONMSS NRC Senior Resident Inspector - LaSalle County Station NRC Project Manager, NRR - LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety

Attachment 1 10 CFR 50.54(q)(5) Procedure Change Summary Analysis Page 1 of 6 Attachment 1 10 CFR 50.54(q)(5) Procedure Change Summary Analysis Procedure/Title Exelon Generation Company, LLC (Exelon) is submitting the following Emergency Plan Addendum revision for LaSalle County Station (LaSalle):

EP-AA-1005, Addendum 3, Revision 3, "Emergency Action Levels for LaSalle Station" Description of Procedure EP-AA-1005, Addendum 3 describes the Emergency Action Levels (EALs) implemented at LaSalle for entering Emergency Classification Levels (ECLs).

Description of Changes The changes included in Revision 3 to EP-AA-1005, Addendum 3 are described below.

1. NEI 99-01 Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," provides the guidance below 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.

Page 2 of 6 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 ISFSIs are covered under ICs HU1 and HA1.

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 NRC Safety Evaluation Report, identify the natural phenomena events and accident 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 casks technical specification located in the Certificate of Compliance.

ECL Assignment Attributes: 3.1.1.B LaSalle has installed the HI-STORM 100 Cask System. The governing Technical Specifications (TS) are in Certificate of Compliance (CoC) No. 1014, Amendment Nos. 3 and 8, Revision 1, Appendix A. The current EAL thresholds encompass the casks in use under Amendment 3. The casks in use under Amendment 8, Revision1 thresholds are being added.

Section 5.7.4 of the TS has the following allowable peak dose rates for a HI-STORM 100 Cask System loaded overpack governed by CoC No. 1014 Amendment No. 8, Revision 1, Appendix A:

30 mrem/hr (gamma + neutron) on the top of the overpack

300 mrem/hr (gamma + neutron) on the side of the overpack excluding inlet and outlet ducts The guidance in NEI 99-01, Revision 6 specifies the use of 2-times the allowable radiation level as the EAL threshold basis. LaSalle EAL E-HU1 was revised as follows to include thresholds for the use of the HI-STORM 100 Cask System governed by CoC No. 1014 Amendment No. 8 Revision 1 Appendix A:

60 mrem/hr (gamma + neutron) on the top of the spent fuel cask

600 mrem/hr (gamma + neutron) on the side of the spent fuel cask excluding inlet and outlet ducts The EAL E-HU1 format was also revised to provide the threshold values in table format as well as revising the units from mr/hr to mrem/hr.

Page 3 of 6 EAL E-HU1 basis refers to the use of the word "cask" versus "overpack" as follows:

"The word cask, as used in this EAL, refers to the storage container in use at the site for dry storage of irradiated fuel."

Revising the wording allows for common terminology to be used in the EAL threshold across the Exelon fleet.

The Basis Reference for EAL E-HU1 was also updated to reference the applicable Amendments of the TS for LaSalle.

2. LaSalle performed an upgrade to their seismic instrumentation in accordance with a site-approved plant design change. This upgrade will change how the seismic instrumentation detects and alarms for an Operating Basis Earthquake (OBE).

The guidance in NEI 99-01, Revision 6, related to EAL HU4.1 specifies the following:

"Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed. Indications available outside the Control Room and/or which require lengthy times to assess (e.g.,

processing of scratch plates or recorded data) should not be used."

Currently, the Initiating Condition (IC) for LaSalle EAL HU4.1 was determined by the following:

"Seismic event > Operating Basis Earthquake (OBE) as indicated by ANY OBE/SSE alarm light/seismic switch alarm relay energized on panel 0PA11J."

In connection with the License Amendment Request (LAR) for LaSalle that requested adoption of the NEI 99-01, Revision 6 EAL scheme changes, there was a question identified in a U.S. Nuclear Regulatory Commission (NRC) preliminary Request for Additional Information (RAI) dated November 11, 2014. In the preliminary RAI under Question No. 9, the NRC requested additional clarification to ensure that the specified IC threshold would result in a timely declaration. Exelon responded to the preliminary RAI by letter dated March 2, 2015, which included the requested information in response to Question No. 9 as follows:

"The seismic panel is located in the Aux Electric Equipment Room located two floors below the Main Control Room. There is an OBE light located on the panel readily available providing indication of > OBE seismic event. It will take approximately 5 min to get to the panel and assess the condition."

The NRC approved the LAR for LaSalle as documented in letter and supporting Safety Evaluation Report (SER) dated July 28, 2015.

The upgraded seismic instrumentation system will provide indication of a seismic event that exceeds OBE in the Main Control Room (MCR) and locally at panel 0PA11J. Per the site-approved plant design change, under the "Alarms and Indication" section of the change package, the indication of an OBE seismic event is as follows:

Page 4 of 6 "The current seismic monitoring system contains a 'Seismic OBE/SSE Level Exceeded' alarm and a Strong Motion Seismic Instr. System Initiated" alarm on annunciator 1ULAN017at MCR panel 1PM10J.

The 'Seismic OBE/SSE Level Exceeded' alarm will currently illuminate when the central recorder has determined that any of the accelerometers has detected movement that exceeds either the OBE or SSE design spectra. The description of this tile will not be modified by this EC, but the tile will only illuminate if the Network Control Center (NCC) has determined that the free field accelerometer has detected movement that exceeds the OBE response spectra and CAV. If the NCC has determined that the free field equipment is not functioning correctly, it will illuminate the MCR room alarm if data from the Reactor Building foundation has exceeded the OBE response spectra."

And "AEER (Aux Electric Equipment Room) panel 0PA11J currently contains annunciator panel lights for Operating Basis Earthquake, Safe Shutdown Earthquake, Seismic System Triggered, and Seismic Switch Alarm Relay. These annunciator lights will be removed. LEDs on the new NCC will replace these alarm tiles. The new NCC is provided with LEDs for the following conditions:

Error

Pooling

Run

Power

Separate warning lights for all five seismic loops

Separate error light for all five seismic loops

Separate trigger light for all five seismic loops

Common Trigger

Common OBE

LED Test/Reset" Based on the above information from site-approved plant design change, EAL HU4.1 was revised as follows:

"Seismic event > Operating Basis Earthquake (OBE) as indicated by:

Alarm on Panel 1PM10J B503 SEISMIC OBE/SSE LEVEL EXCEEDED OR

OBE LED lit on 0PA11J The first bulleted threshold aligns with the guidance in NEI 99-01, Revision 6, since this indication is immediately available to MCR personnel and can be readily assessed. The second bulleted threshold was previously assessed as being timely in Exelon's response to the RAI (i.e., Question No. 9) submitted by letter dated March 2, 2015, and subsequently approved by the NRC in its letter and supporting SER dated July 28, 2015.

Page 5 of 6 Description of How the Changes Still Comply with Regulations

1. The change to the EAL E-HU1 reflects the use of an additional type of HI-STORM 100 Cask System Overpack at the ISFSI at LaSalle. The CoC No. 1014, Amendment No. 8, Revision 1, Appendix A documents the allowable peak dose rates for the additional type of HI-STORM 100 Cask System Overpack.

NEI 99-01, Rev 6 provides the guidance for the development of the EAL E-HU1 threshold as 2 x the allowable peak dose rates. This adds additional thresholds 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. Also the formatting is revised to provide threshold values in table format and units are revised from mr/hr to mrem/hr to be consistent with the units used in LaSalle Station Certificate of Compliance Number 1014, Amendment No. 8 Revision 1 Appendix A.

An additional change to the E-HU1 Basis Reference was made to include LaSalle Station Certificate of Compliance Number 1014, Amendment No 8 Revision 1 Appendix A.

Updating the EAL threshold values based on an approved Certificate Of Compliance does not alter the meaning or intent of the basis of the approved EAL. The addition of the new type of ISFSI Cask was reviewed and approved in the Safety Evaluation Report Docket No.

72-1014 Holtec International HI-STORM 100 Cask System Certificate of Compliance No.

1014 Amendment No. 8, Revision No.1. This change is not a Reduction In Effectiveness because no existing EP requirements have been deleted or minimized under this revision.

Revising the EAL to provide threshold values in table format as well as updating the reference section and revising the units from mr/hr to mrem/hr do not alter the meaning or intent of the approved EAL. The regulations and commitments to the NRC continue to be met.

Regulatory Guide (RG) 1.219, "Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors," provides the following guidance related to adding the new threshold for the approved ISFSI cask:

f. The following examples would generally not require prior NRC approval: (1) A change to an EAL numeric threshold to reflect an approved change in a technical specification, provided that the basis of the approved EAL is unchanged (e.g., an EAL basis refers to a particular technical specification but not a limiting condition for operation value),

2. The revision to LaSalle EAL HU4.1 reflects the upgrades to the seismic instrumentation in accordance with a site-approve plant design change.

NEI 99-01, Revision 6 provides the guidance for the development of EAL HU4.1, and states in part:

"Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed."

The NRC's acceptance and approval of the Exelon's NEI 99-01, Revision 6 EAL scheme change LAR as documented in a letter and supporting SER dated July 28, 2015, ensures the ability for timely classification if using indications provided for at panel 0PA11J.

Page 6 of 6 The revised threshold uses both MCR indication and indication at panel 0PA11J for making the greater than OBE seismic event.

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

1. Updating the EAL threshold values based on an approved CoC does not alter the meaning or intent of the basis of the approved EAL. The addition of the new type of ISFSI Cask System was reviewed and approved in the Safety Evaluation Report Docket No. 72-1014 Holtec International HI-STORM 100 Cask System CoC No. 1014 Amendment No. 8, Revision 1. Revising the EAL to provide threshold values in table format as well as updating the reference section and revising the units from mr/hr to mrem/hr do not alter the meaning or intent of the approved EAL. The applicable regulations and commitments to the NRC continue to be met.

RG 1.219 provides the following guidance related to adding the new threshold for the approved ISFSI cask:

f. The following examples would generally not require prior NRC approval: (1) A change to an EAL numeric threshold to reflect an approved change in a technical specification, provided that the basis of the approved EAL is unchanged (e.g., an EAL basis refers to a particular technical specification but not a limiting condition for operation value),

2. Updating the EAL threshold to conform with the upgraded seismic monitoring system and basing the threshold on NRC-approved guidance contained in NEI 99-01, Revision 6 and an approved license amendment does not alter the meaning or intent of the basis of the approved EAL. The applicable regulations and commitments to the NRC continue to be met.

Updating the EAL threshold values based on an approved CoC and a site-approved plant design change for the upgrading seismic monitoring system does not alter the meaning or intent of the basis of the approved EAL. Emergency planning commitments to the NRC continue to be met and no emergency planning requirements have been deleted or minimized under this revision. The changes are also supported by the guidance in RG 1.219 as noted above related to adding a new threshold for the approved ISFSI cask.

Therefore, the changes described do not result in a reduction in effectiveness of the Emergency Plan for LaSalle.

Attachment 2 EP-AA-1005, Addendum 3, Revision 3, "Emergency Action Levels for LaSalle Station" Emergency Plan Addendum Revision

EP-AA-1005 Addendum 3 Revision 3 EXELON NUCLEAR EMERGENCY ACTION LEVELS FOR LASALLE STATION

LaSalle Annex Exelon Nuclear REVISION HISTORY Rev. 0 December 2014 Rev. 1 January 2016 Rev. 2 September 2016 Rev. 3 October 2017 October 2017 i EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 first 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.

October 2017 LS 1-1 EP-AA-1005 Addendum 3(Revision 3)

LaSalle 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-AA-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.

Individuals responsible for the classification of events will refer to the Initiating Condition 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.

October 2017 LS 1-2 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear 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:

EAL

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 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.

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 Station Emergency Director (Technical Support Center).

Classification of emergencies remains the responsibility of the applicable onsite facility even after Command and Control is transferred to the Corporate Emergency Director (Emergency Operations Facility).

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.

October 2017 LS 1-3 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear 1.2 Classification, Instrumentation and Transient Events Classifications are based on evaluation of each Unit. 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 indications operability, the conditions existence, or the reports 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).

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 sites Technical Specifications.

Activities which cause the site to operate beyond that allowed by the sites 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.

October 2017 LS 1-4 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear When two or more EALs are determined, declaration will be made on the highest classification level for the Unit. When both units are affected, the highest classification for the Station will be used for notification purposes and both Units ECLs will be noted.

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.

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).

October 2017 LS 1-5 EP-AA-1005 Addendum 3(Revision 3)

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

FPB 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 any FPB thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the FPB 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 any 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 FPB 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.

The reestablishment of a FPB does not alter or lower the existing classification.

Termination and entry into RECOVERY phase is still required for exiting the October 2017 LS 1-6 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear present classification. However the reestablishment of the barrier should be considered in determining future classifications should plant conditions or events change.

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

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.

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.

FISSION PRODUCT BARRIER (FPB) THRESHOLD: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

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.

October 2017 LS 1-7 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear 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.

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 > 200° F (4) Cold Shutdown (a): Shutdown 200° 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)

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.

October 2017 LS 1-8 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear 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 component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

October 2017 LS 1-9 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex Exelon Nuclear Emergency Action Level Technical Basis Page Index General Site Area Alert Unusual Event EAL Pg. EAL Pg. EAL Pg. EAL Pg.

RG1 2-25 RS1 2-27 RA1 2-29 RU1 2-32 RG2 2-35 RS2 2-36 RA2 2-37 RU2 2-39 RA3 2-41 RU3 2-44 FG1 2-45 FS1 2-46 FA1 2-47 Fuel Clad RCS Containment FC1 2-48 FC2 2-49 RC2 2-53 CT2 2-60 RC3 2-55 CT3 2-61 RC4 2-56 FC5 2-51 RC5 2-58 CT5 2-63 CT6 2-64 FC7 2-52 RC7 2-59 CT7 2-67 MG1 2-68 MS1 2-70 MA1 2-71 MU1 2-73 MG2 2-74 MS2 2-76 MS3 2-77 MA3 2-79 MU3 2-81 MA4 2-84 MU4 2-86 MA5 2-88 MU6 2-90 MU7 2-92 CA1 2-94 CU1 2-96 CA2 2-98 CU3 2-100 CU4 2-101 CA5 2-103 CU5 2-105 CG6 2-107 CS6 2-110 CA6 2-113 CU6 2-115 HG1 2-118 HS1 2-120 HA1 2-122 HU1 2-125 HS2 2-127 HA2 2-129 HU3 2-130 HU4 2-134 HA5 2-136 HU6 2-139 HG7 2-142 HS7 2-143 HA7 2-144 HU7 2-145 E-HU1 2-146 October 2017 LS 1-10 EP-AA-1005 Addendum 3(Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluents RG1 Release of gaseous radioactivity 1 2 3 4 5 D RS1 Release of gaseous radioactivity 1 2 3 4 5 D RA1 Release of gaseous or liquid 1 2 3 4 5 D RU1 Release of gaseous or liquid 1 2 3 4 5 D resulting in offsite dose greater than 1,000 mRem TEDE resulting in offsite dose greater than 100 mRem TEDE radioactivity resulting in offsite dose greater than radioactivity greater than or 5,000 mRem thyroid CDE. or 500 mRem thyroid CDE. 10 mrem TEDE or 50 mrem thyroid CDE. 2 times the ODCM limits for 60 minutes or longer.

Emergency Action Level (EAL):

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

Notes: Notes: Notes:

The Emergency Director should declare the event promptly

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

The Emergency Director should declare the event

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

If an ongoing release is detected and the release start time

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

If an ongoing release is detected and the release start

If an ongoing release is detected and the release start is unknown, assume that the release duration has exceeded 15 minutes. time is unknown, assume that the release duration has time is unknown, assume that the release duration has exceeded 15 minutes. exceeded 15 minutes. exceeded 60 minutes.

Classification based on effluent monitor readings assumes

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

Classification based on effluent monitor readings assumes

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

The pre-calculated effluent monitor values presented in

The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification

The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until dose assessment results are available. EAL #1 should be used for emergency classification assessments until dose assessment results are available. assessments until dose assessment results are available. 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for 60 minutes.

Radiological Effluents

1. The sum of readings on the Vent Stack and SBGT WRGMs > 1. The sum of readings on the Vent Stack and SBGT WRGMs > 1. The sum of readings on the Vent Stack and SBGT WRGMs >

8.26E+09 uCi/sec for > 15 minutes (as determined by Control 8.26E+08 uCi/sec for > 15 minutes (as determined by 8.26E+07 uCi/sec for > 15 minutes (as determined by

Radwaste Effluent Monitor LCRM 0D18-K606 Room Panels or PPDS - Total Noble Gas Release Rate). Control Room Panels or PPDS - Total Noble Gas Release Control Room Panels or PPDS - Total Noble Gas Release Rate). OR OR Rate).

OR OR

Discharge Permit specified monitor

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 2. Dose assessment using actual meteorology indicates doses at OR

a. > 1000 mRem TEDE or beyond the site boundary of EITHER: or beyond the site boundary of EITHER:

2. The sum of readings on the Vent Stack and SBGT WRGMs >

OR a. > 10 mRem TEDE 9.42 E+05 uCi/sec for > 60 minutes (as determined by

a. > 100 mRem TEDE

b. > 5000 mRem CDE Thyroid OR Control Room Panels or PPDS - Total Noble Gas Release OR Rate).

OR b. > 50 mRem CDE Thyroid

3. Field survey results at or beyond the site boundary indicate b. > 500 mRem CDE Thyroid OR OR EITHER: OR 3. Analysis of a liquid effluent sample indicates a concentration 3. Confirmed sample analyses for gaseous or liquid releases

a. Gamma (closed window) dose rates or release rate that would result in doses greater than indicate concentrations or release rates

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

>1000 mR/hr are expected to continue for EITHER: > 2 times ODCM Limit with a release duration of

> 60 minutes. a. 10 mRem TEDE for 60 minutes of exposure > 60 minutes.

a. Gamma (closed window) dose rates OR OR

>100 mR/hr are expected to continue for b. 50 mRem CDE Thyroid for 60 minutes of

b. Analyses of field survey samples indicate

> 60 minutes. exposure

> 5000 mRem CDE Thyroid for 60 minutes of inhalation. 4. Field survey results at or beyond the site boundary indicate OR EITHER:

b. Analyses of field survey samples indicate a. Gamma (closed window) dose rates

> 500 mRem CDE Thyroid for 60 minutes of > 10 mR/hr are expected to continue for > 60 inhalation. 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 October 2017 HOT MATRIX LS 2-1 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluents RG2 Spent fuel pool level cannot be 1 2 3 4 5 D RS2 Spent fuel pool level at 1 2 3 4 5 D RA2 Significant lowering of water 1 2 3 4 5 D RU2 UNPLANNED loss of water level 1 2 3 4 5 D restored to at least 820.42 ft. as indicated on 1(2) LI- 820.42 ft. as indicated on 1(2) LI-FC165. level above, or damage to, irradiated fuel. above irradiated fuel FC165 for 60 minutes or longer.

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

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

1. a. UNPLANNED water level drop in the REFUELING Note: The Emergency Director should declare the General Lowering of spent fuel pool level to 820.42 ft. as indicated on 1. Uncovery of irradiated fuel in the REFUELING PATHWAY as indicated by ANY of the following:

Emergency promptly upon determining that the 1(2) LI-FC165. PATHWAY.

applicable time has been exceeded, or will likely be

Refueling Cavity water level < 340 in.

OR exceeded. on shutdown range.

2. Damage to irradiated fuel resulting in a release of Spent fuel pool level cannot be restored to at least radioactivity from the fuel as indicated by ARM OR 820.42 ft. as indicated on 1(2) LI-FC165 for 60 minutes or 0D21-K604A Radiation Monitor reading

Spent Fuel Pool water level < 21 ft. 4 longer. >1000 mRem/hr. in.

Table R2 OR OR Areas Requiring Continuous Occupancy 3. Lowering of spent fuel pool level to 830.00 ft. as

Indication or report of a drop in water indicated on 1(2) LI-FC165. level in the REFUELING PATHWAY.

Main Control Room (1(2)D18-K751A-D)

Radiological Effluents AND

Central Alarm Station - (by survey) b. UNPLANNED Area Radiation Monitor reading rise on ARM 0D21-K604A radiation monitor.

RA3 Radiation levels that impede 1 2 3 4 5 D RU3 Reactor coolant activity greater 1 2 3 access to equipment necessary for normal plant than Technical Specification allowable limits.

Table R3 operations, cooldown or shutdown.

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

Emergency Action Level (EAL):

Entry Related Mode Note: If the equipment in the listed room or area was Area Applicability already inoperable, or out of service, before the 1 Offgas system radiation monitor HI-HI alarm.

Reactor Building event occurred, then no emergency classification is 673 SE Corner Room warranted OR 694 NW Corner Room 2. Specific coolant activity > 4.0 uCI/gm Dose SE Corner Room 1. Dose rate > 15 mR/hr in ANY of the areas equivalent I-131.

Raceway contained in Table R2.

Mode 3 and 4 740 NE Valve Room OR NE 2. UNPLANNED event results in radiation levels that SW prohibit or significantly impede access to ANY of the 761 NE following Table R3 plant rooms or areas.

SW Auxiliary Building 674 CSCS Rooms Mode 3 and 4 731 AEER 736 DG Penthouse Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-2 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT FG1 Loss of any two barriers AND Loss or Potential Loss of third barrier. 1 2 3 FS1 Loss or Potential Loss of ANY two barriers. 1 2 3 FA1 ANY Loss or ANY Potential Loss of either Fuel Clad or RCS 1 2 3 FC - Fuel Clad RC - Reactor Coolant System CT - Containment Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Coolant activity > 300 uCi/gm Dose None None

1. RCS Activity Equivalent I-131. None None None

2. RPV water level cannot be restored and 1. RPV water level cannot be restored and

1. Plant conditions indicate Primary maintained > -161 inches (TAF) maintained > -161 inches (TAF) Plant conditions indicate Primary Containment

2. RPV Water Containment flooding is required. OR OR None None flooding is required.

Level

3. RPV water level cannot be determined. 2. RPV water level cannot be determined.

3. Drywell pressure > 45 psig and rising.

OR

1. UNPLANNED rapid drop in Drywell pressure 4. a. Drywell or suppression chamber hydrogen

1. Drywell pressure >1.77 psig. concentration > 6%.

following Drywell pressure rise.

3. Primary AND AND OR Containment None None 2. Drywell pressure rise is due to RCS None b. Drywell or suppression chamber oxygen

2. Drywell pressure response not consistent Pressure/Conditions leakage concentration > 5%.

with LOCA conditions.

OR

5. Heat Capacity Limit (LGA-003, Fig.H) exceeded.

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

1. UNISOLABLE Main Steam Line (MSL), a. Secondary Containment area RCIC, Feedwater, or RWCU line break. temperature > LGA-002 Maximum OR Normal operating levels.

4.RCS Leak Rate None None None None

2. Emergency RPV Depressurization is OR required. b. Secondary Containment area radiation level > LGA-002 Maximum Normal operating level.

5.Primary Drywell radiation monitor reading Drywell radiation monitor reading Drywell radiation monitor reading Containment > 1.9 E+2 R/hr (> 190 R/hr). None > 1.00E+2 R/hr (>100 R/hr). None None > 4.35 E+02 R/hr (> 435 R/hr).

Radiation

1. UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

OR

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

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

a. Secondary Containment area temperature > LGA-002, Maximum Safe operating levels.

OR

b. Secondary Containment area radiation level > LGA-002, Maximum Safe operating levels.

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

7. Emergency Emergency Director that indicates Loss Emergency Director that indicates Potential Emergency Director that indicates Loss of Emergency Director that indicates Potential Emergency Director that indicates Loss of the Emergency Director that indicates Potential Loss Director Judgment of the Fuel Clad Barrier. Loss of the Fuel Clad Barrier. the RCS Barrier. Loss of the RCS Barrier. Containment Barrier. of the Containment Barrier.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-3 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MG1 Prolonged loss of all offsite 1 2 3 MS1 Loss of all Off-site and On-Site 1 2 3 MA1 Loss of all but one AC power 1 2 3 MU1 Loss of all offsite AC power 1 2 3 and all onsite AC power to emergency buses. AC power to emergency busses for 15 minutes or source to emergency buses for 15 minutes or capability to emergency buses for 15 minutes or longer. longer. longer.

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

Emergency 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 Loss of AC Power 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.

1. Loss of ALL offsite AC power to unit ECCS busses 1. AC power capability to unit ECCS busses (excluding (excluding Division 3). Division 3) reduced to only one of the following power Loss of ALL offsite AC power capability to unit ECCS

1. Loss of ALL offsite AC Power to unit ECCS busses AND sources for > 15 minutes. busses (excluding Division 3) for > 15 minutes.

(excluding Division 3).

2. Failure of DG 0 and DG 1A(2A) emergency diesel AND

System Auxiliary Transformer 142(242) generators to supply power to unit ECCS busses.

2. Failure of DG 0 and DG 1A(2A) emergency diesel

Unit Auxiliary Transformer 141(241)

AND generators to supply power to unit ECCS busses.

Unit Emergency Diesel Generator DG 1A(2A)

3. EITHER of the following: AND

Shared Emergency Diesel Generator DG 0

a. Restoration of at least one unit ECCS bus (excluding 3. Failure to restore power to at least one unit ECCS bus Division 3) in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely.

Other SAT via crosstie breakers (excluding Division 3) in < 15 minutes from the time of OR loss of both offsite and onsite AC power. AND

b. RPV water level cannot be restored and maintained 2. ANY additional single power source failure (excluding

> -150 inches. Division 3) will result in a loss of ALL AC power to SAFETY SYSTEMS.

MG2 Loss of all AC and Vital DC power sources 1 2 3 MS2 Loss of all Vital DC power for 1 2 3 for 15 minutes or longer. 15 minutes or longer.

Emergency Action Level (EAL):

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

has been exceeded, or will likely be exceeded.

1. Loss of ALL offsite AC power to unit ECCS busses (excluding Division 3).

Voltage is < 108 VDC on 125 VDC battery busses AND 111Y(211Y) and 112Y(212Y) for >15 minutes.

2. Failure of DG 0 and DG 1A(2A) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Voltage is < 108 VDC on unit 125 VDC battery busses 111Y(211Y) and 112Y(212Y).

AND

4. ALL AC and Vital DC power sources (excluding Division 3) have been lost for > 15 minutes.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-4 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MS3 Inability to shutdown the reactor 1 2 MA3 Automatic or manual trip fails 1 2 MU3 Automatic or manual trip fails 1 2 causing a challenge to RPV water level or RCS heat to shutdown the reactor, and subsequent to shutdown the reactor.

removal. manual actions taken at the reactor control consoles are not successful in shutting down Emergency Action Level (EAL):

the reactor. Note: A manual action is any operator action, or set of Emergency Action Level (EAL): actions, which causes the control rods to be rapidly Emergency Action Level (EAL): inserted into the core, and does not include manually

1. Automatic scram did not shutdown the reactor as indicated by Reactor Power > 3%. Note: A manual action is any operator action, or set of driving in control rods or implementation of boron actions, which causes the control rods to be rapidly injection strategies.

AND inserted into the core, and does not include manually 1. a. Automatic scram did not shutdown the reactor as

2. ALL manual / ARI actions to shutdown the reactor have driving in control rods or implementation of boron indicated by Reactor Power > 3%.

been unsuccessful as indicated by Reactor Power > 3%. injection strategies.

AND RPS Failure AND 1. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 3%. b. Subsequent manual / ARI action taken at the

3. EITHER of the following conditions exist: Reactor Console is successful in shutting down AND the reactor as indicated by Reactor Power < 3%.

RPV water level cannot be restored and maintained

> -150 inches on WR (-183 inches (MSCRWL) on FZ 2. Manual / ARI actions taken at the Reactor Console are OR if WR not available) not successful in shutting down the reactor as indicated by Reactor Power > 3%. 2. a. Manual scram did not shutdown the reactor as OR indicated by Reactor Power > 3%.

Heat Capacity Limit (LGA-003, Fig. H) exceeded. AND

b. EITHER of the following:

1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor as indicated by Reactor Power < 3%.

OR

2. Subsequent automatic scram / ARI is successful in shutting down the reactor as indicated by Reactor Power < 3%.

MA4 UNPLANNED loss of Control Room 1 2 3 MU4 UNPLANNED loss of Control Room 1 2 3 indications for 15 minutes or longer with a significant indications for 15 minutes or longer.

Control Room Indications transient in progress.

Table M2 Significant Transients Emergency Action Level (EAL): Emergency Action Level (EAL):

Table M1 Control Room Parameters Note: The Emergency Director should declare the event Note: The Emergency Director should declare the event

Turbine Trip promptly upon determining that the applicable time promptly upon determining that the applicable time

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

RPV Water Level

Reactor Scram

RPV Pressure

ECCS Activation 1. UNPLANNED event results in the inability to monitor UNPLANNED event results in the inability to monitor ANY

Primary Containment Pressure

Recirc. Runback > 25% Reactor Power ANY Table M1 parameters from within the Control Table M1 parameters from within the Control Room for > 15

Suppression Pool Level Change Room for > 15 minutes. minutes.

Suppression Pool Temperature AND

Thermal Power oscillations > 10% Reactor Power Change 2. Any Table M2 transient in progress.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-5 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MA5 Hazardous event affecting a 1 2 3 SAFETY SYSTEM required for the current operating mode.

Emergency Action Level (EAL):

Note: If it is determined that the conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.

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

Seismic event (earthquake)

Internal or external flooding event

High winds or tornado strike Hazard affects Safety System

FIRE

EXPLOSION

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

2. EITHER of the following:

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

OR

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

MU6 RCS leakage for 15 minutes 1 2 3 or longer.

Emergency Action Level (EAL):

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

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-6 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT System Malfunction MU7 Loss of all On-site or Off-site 1 2 3 Table M3 Communications Capability communication capabilities.

System Onsite Offsite NRC Emergency Action Level (EAL):

Communications Plant Radio X Plant Page X 1. Loss of ALL Table M3 Onsite communications capability affecting the ability to perform routine All telephone Lines X X X operations.

(Commercial and microwave) OR ENS X X 2. Loss of ALL Table M3 Offsite communication capability affecting the ability to perform offsite HPN X X notifications.

Satellite Phones X X OR

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

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-7 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HG1 HOSTILE ACTION resulting in loss 1 2 3 4 5 D HS1 HOSTILE ACTION within the 1 2 3 4 5 D HA1 HOSTILE ACTION within the 1 2 3 4 5 D HU1 Confirmed SECURITY CONDITION 1 2 3 4 5 D of physical control of the facility PROTECTED AREA OWNER CONTROLLED AREA or airborne attack or threat.

threat within 30 minutes.

Emergency Action Level (EAL):

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

Hostile Action

1. A notification from the Security Force that a HOSTILE 1. Notification of a credible security threat directed at the A notification from the Security Force that a HOSTILE 1. A validated notification from NRC from an aircraft ACTION is occurring or has occurred within the site as determined per SY-AA-101-132, Security ACTION is occurring or has occurred within the attack threat < 30 minutes of the site.

PROTECTED AREA. Assessment and Response to Unusual Activities.

PROTECTED AREA.

AND OR OR

2. a. ANY Table H1 safety function cannot be 2. Notification by the Security Force that a HOSTILE 2. A validated notification from the NRC providing controlled or maintained. ACTION is occurring or has occurred within the information of an aircraft threat.

OWNER CONTROLED AREA. OR OR

3. Notification by the Security Force of a SECURITY

b. Damage to spent fuel has occurred or is CONDITION that does not involve a HOSTILE IMMINENT ACTION.

HS2 Inability to control a key safety 1 2 3 4 5 D HA2 Control Room evacuation resulting 1 2 3 4 5 D function from outside the Control Room in transfer of plant control to alternate locations Table H1 Safety Functions Emergency Action Level (EAL): Emergency Action Level (EAL):

Reactivity Control Note: The Emergency Director should declare the event A Control Room evacuation has resulted in plant control (ability to shutdown the reactor and keep it shutdown) promptly upon determining that the applicable time has being transferred from the Control Room to alternate Transfer of Plant Control been exceeded, or will likely be exceeded. locations per:

RPV Water Level (ability to cool the core)

1. A Control Room evacuation has resulted in plant control

LOA-RX-101(201), Unit 1(2) Control Room being transferred from the Control Room to alternate Evacuation Abnormal

RCS Heat Removal (ability to maintain a heatsink) locations per: OR

LOA-RX-101(201), Unit 1(2) Control Room

LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Evacuation Abnormal Fire in the Control Room OR AEER.

OR

LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Fire in the Control Room OR AEER.

AND

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 October 2017 HOT MATRIX LS 2-8 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HU3 FIRE potentially degrading the level 1 2 3 4 5 D of safety of the plant.

Emergency Action Level (EAL):

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

Escalation of the emergency classification level would be via IC CA2 or MA5 Table H2 Vital Areas

Reactor Building 1. A FIRE in ANY Table H2 area is not extinguished in (when inerted the Drywell is exempt) < 15-minutes of ANY of the following FIRE detection

Control Room indications:

Fire

Auxiliary Building

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

Receipt of multiple (more than 1) fire alarms or

Unit and Shared Emergency Diesel indications Generator Rooms

Field verification of a single fire alarm

Switchgear and Battery Rooms OR

Remote Shutdown Rooms 2. a. Receipt of a single fire alarm in ANY Table H2 area

CSCS Pump Rooms (i.e., no other indications of a FIRE).

LSH (for 0E12-F300 access only) 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 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-9 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HU4 Seismic event greater than OBE levels 1 2 3 4 5 D Emergency Action Leve l (EAL):

Note: Escalation of the emergency classification level would be via IC CA2 or MA5 For emergency classification if EAL # 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in < 15 mins of the event.

1. Seismic event > Operating Basis Earthquake (OBE) as indicated by:

Earthquake

Alarm on Panel 1PM10J B503 SEISMIC OBE/SSE LEVEL EXCEEDED OR

OBE LED lit on 0PA11J 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 mins of the event:

The earthquake resulted in Modified Mercalli Intensity (MMI) > VI and occurred < 3.5 miles of the plant.

The earthquake was magnitude > 6.0

The earthquake was magnitude > 5.0 and occurred

< 125 miles of the plant.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-10 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HA5 Gaseous release impeding access to 3 4 Table H3 equipment necessary for normal plant operations, Areas with Entry Related Mode Applicability cooldown or shutdown.

Entry Related Mode Area Applicability Emergency Action Level (EAL):

Reactor Building Note: If the equipment in the listed room or area was Toxic Gas 673 SE Corner Room already inoperable, or out of service, before the 694 NW Corner Room event occurred, then no emergency classification SE Corner Room is warranted.

Raceway Mode 3 and 4 1. Release of a toxic, corrosive, asphyxiant or 740 NE Valve Room NE flammable gas in ANY Table H3 area.

SW AND 761 NE SW 2. Entry into the room or area is prohibited or impeded Auxiliary Building 674 CSCS Rooms Mode 3 and 4 731 AEER 736 DG Penthouse HU6 Hazardous Event 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 Hazardous Event

1. Tornado strike within the PROTECTED AREA.

OR

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

OR

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

OR

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

October 2017 HOT MATRIX LS 2-11 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HG7 Other conditions exist which in the 1 2 3 4 5 D HS7 Other conditions exist which in the 1 2 3 4 5 D HA7 Other conditions exist which in the 1 2 3 4 5 D HU7 Other conditions exist which in the 1 2 3 4 5 D judgment of the Emergency Director warrant judgment of the Emergency Director warrant judgment of the Emergency Director warrant judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY. declaration of a SITE AREA EMERGENCY. declaration of an ALERT. declaration of an UNUSUAL 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 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 Emergency Director, indicate that events are in progress or 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 have occurred which involve an actual or potential 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 substantial degradation of the level of safety of the plant or a 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, security event that involves probable life threatening risk to 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 site personnel or damage to site equipment because of 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 HOSTILE ACTION. Any releases are expected to be limited degradation of safety systems occurs.

Guideline exposure levels offsite for more than the immediate equipment needed for the protection of the public. Any to small fractions of the EPA Protective Action Guideline site area. releases are not expected to result in exposure levels which exposure levels.

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

Emergency Director Judgment Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-12 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex HOT MATRIX HOT MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 Damage to a loaded cask 1 2 3 4 5 D CONFINEMENT BOUNDARY.

Emergency Action Level (EAL):

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by exceeding ANY Table E1 on-contact radiation reading.

Table E1 On-contact radiation readings ISFSI Location of on-contact radiation Cask Type reading Top of the Side of the spent fuel spent fuel cask, excluding inlet HI-STORM cask and outlet ducts (mrem/hr) (mrem/hr)

Labeled as

> 40 > 220 XXX-A3 Labeled as XXX-A8 > 60 > 600 Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 HOT MATRIX LS 2-13 HOT MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluents RG1 Release of gaseous radioactivity 1 2 3 4 5 D RS1 Release of gaseous radioactivity 1 2 3 4 5 D RA1 Release of gaseous or liquid 1 2 3 4 5 D RU1 Release of gaseous or liquid 1 2 3 4 5 D resulting in offsite dose greater than 1,000 mRem TEDE resulting in offsite dose greater than 100 mRem TEDE radioactivity resulting in offsite dose greater than radioactivity greater than or 5,000 mRem thyroid CDE. or 500 mRem thyroid CDE. 10 mrem TEDE or 50 mrem thyroid CDE. 2 times the ODCM limits for 60 minutes or longer.

Emergency Action Level (EAL):

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

Notes: Notes: Notes:

The Emergency Director should declare the event promptly

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

The Emergency Director should declare the event

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

If an ongoing release is detected and the release start time

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

If an ongoing release is detected and the release start

If an ongoing release is detected and the release start is unknown, assume that the release duration has exceeded 15 minutes. time is unknown, assume that the release duration has time is unknown, assume that the release duration has exceeded 15 minutes. exceeded 15 minutes. exceeded 60 minutes.

Classification based on effluent monitor readings assumes

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

Classification based on effluent monitor readings assumes

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

The pre-calculated effluent monitor values presented in

The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification

The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until dose assessment results are available. EAL #1 should be used for emergency classification assessments until dose assessment results are available. assessments until dose assessment results are available. 1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for 60 minutes.

Radiological Effluents

1. The sum of readings on the Vent Stack and SBGT WRGMs > 1. The sum of readings on the Vent Stack and SBGT WRGMs > 1. The sum of readings on the Vent Stack and SBGT WRGMs >

8.26E+09 uCi/sec for > 15 minutes (as determined by Control 8.26E+08 uCi/sec for > 15 minutes (as determined by 8.26E+07 uCi/sec for > 15 minutes (as determined by

Radwaste Effluent Monitor LCRM 0D18-K606 Room Panels or PPDS - Total Noble Gas Release Rate). Control Room Panels or PPDS - Total Noble Gas Release Control Room Panels or PPDS - Total Noble Gas Release Rate). OR OR Rate).

OR OR

Discharge Permit specified monitor

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 2. Dose assessment using actual meteorology indicates doses at OR

a. > 1000 mRem TEDE or beyond the site boundary of EITHER: or beyond the site boundary of EITHER:

2. The sum of readings on the Vent Stack and SBGT WRGMs >

OR a. > 10 mRem TEDE 9.42 E+05 uCi/sec for > 60 minutes (as determined by

a. > 100 mRem TEDE

b. > 5000 mRem CDE Thyroid OR Control Room Panels or PPDS - Total Noble Gas Release OR Rate).

OR b. > 50 mRem CDE Thyroid

3. Field survey results at or beyond the site boundary indicate b. > 500 mRem CDE Thyroid OR OR EITHER: OR 3. Analysis of a liquid effluent sample indicates a concentration 3. Confirmed sample analyses for gaseous or liquid releases

a. Gamma (closed window) dose rates or release rate that would result in doses greater than indicate concentrations or release rates

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

>1000 mR/hr are expected to continue for EITHER: > 2 times ODCM Limit with a release duration of

> 60 minutes. a. 10 mRem TEDE for 60 minutes of exposure > 60 minutes.

a. Gamma (closed window) dose rates OR OR

>100 mR/hr are expected to continue for b. 50 mRem CDE Thyroid for 60 minutes of

b. Analyses of field survey samples indicate

> 60 minutes. exposure

> 5000 mRem CDE Thyroid for 60 minutes of inhalation. 4. Field survey results at or beyond the site boundary indicate OR EITHER:

b. Analyses of field survey samples indicate a. Gamma (closed window) dose rates

> 500 mRem CDE Thyroid for 60 minutes of > 10 mR/hr are expected to continue for > 60 inhalation. 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 October 2017 COLD MATRIX LS 2-14 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluents RG2 Spent fuel pool level cannot be 1 2 3 4 5 D RS2 Spent fuel pool level at 1 2 3 4 5 D RA2 Significant lowering of water 1 2 3 4 5 D RU2 UNPLANNED loss of water level 1 2 3 4 5 D restored to at least 820.42 ft. as indicated on 1(2) LI- 820.42 ft. as indicated on 1(2) LI-FC165. level above, or damage to, irradiated fuel. above irradiated fuel FC165 for 60 minutes or longer.

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

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

1. a. UNPLANNED water level drop in the REFUELING Note: The Emergency Director should declare the General Lowering of spent fuel pool level to 820.42 ft. as indicated on 1. Uncovery of irradiated fuel in the REFUELING PATHWAY as indicated by ANY of the following:

Emergency promptly upon determining that the 1(2) LI-FC165. PATHWAY.

applicable time has been exceeded, or will likely be

Refueling Cavity water level < 340 in.

OR exceeded. on shutdown range.

2. Damage to irradiated fuel resulting in a release of Spent fuel pool level cannot be restored to at least radioactivity from the fuel as indicated by ARM OR 820.42 ft. as indicated on 1(2) LI-FC165 for 60 minutes or 0D21-K604A Radiation Monitor reading

Spent Fuel Pool water level < 21 ft. 4 longer. >1000 mRem/hr. in.

Table R2 OR OR Areas Requiring Continuous Occupancy 3. Lowering of spent fuel pool level to 830.00 ft. as

Indication or report of a drop in water indicated on 1(2) LI-FC165. level in the REFUELING PATHWAY.

Main Control Room (1(2)D18-K751A-D)

Radiological Effluents AND

Central Alarm Station - (by survey) b. UNPLANNED Area Radiation Monitor reading rise on ARM 0D21-K604A radiation monitor.

RA3 Radiation levels that impede 1 2 3 4 5 D access to equipment necessary for normal plant Table R3 operations, cooldown or shutdown.

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

Entry Related Mode Note: If the equipment in the listed room or area was Area Applicability already inoperable, or out of service, before the Reactor Building event occurred, then no emergency classification is 673 SE Corner Room warranted 694 NW Corner Room SE Corner Room 1. Dose rate > 15 mR/hr in ANY of the areas Raceway contained in Table R2.

Mode 3 and 4 740 NE Valve Room OR NE 2. UNPLANNED event results in radiation levels that SW prohibit or significantly impede access to ANY of the 761 NE following Table R3 plant rooms or areas.

SW Auxiliary Building 674 CSCS Rooms Mode 3 and 4 731 AEER 736 DG Penthouse Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defuel October 2017 COLD MATRIX LS 2-15 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Cold Shutdown / Refueling System Malfunctions CA1 Loss of all offsite and onsite AC power 4 5 D CU1 Loss of all but one AC power source 4 5 D to emergency busses for 15 minutes or longer. to emergency buses for 15 minutes or longer.

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

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

1. Loss of ALL offsite AC power to unit ECCS busses 1. AC power capability to unit ECCS busses (excluding (excluding Division 3). Division 3) reduced to only one of the following power AND sources for > 15 minutes.

System Auxiliary Transformer 142(242)

2. Failure of DG 0 and DG 1A(2A) emergency diesel generators to supply power to unit ECCS busses.

Unit Auxiliary Transformer 141(241)

Unit Emergency Diesel Generator DG 1A(2A)

AND

Shared Emergency Diesel Generator DG 0

3. Failure to restore power to at least one unit ECCS

Other SAT via crosstie breakers bus (excluding Division 3) in < 15 minutes from the AND time of loss of both offsite and onsite AC power.

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

CA2 Hazardous event affecting 4 5 SAFETY SYSTEM required for the current operating mode.

Emergency Action Level (EAL):

Note: If it is determined that the conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.

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

Seismic event (earthquake)

Safety System

Internal or external flooding event

High winds or tornado strike

FIRE

EXPLOSION

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

2. EITHER of the following:

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

OR

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

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-16 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Cold Shutdown / Refueling System Malfunctions CU3 Loss of Vital DC power for 15 minutes 4 5 or longer.

DC Power 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 < 108 VDC on required unit 125 VDC battery busses 111Y(211Y) and 112Y(212Y) for

> 15 minutes.

CU4 Loss of all onsite or offsite communication 4 5 D Table C1 Communications Capability capabilities.

System Onsite Offsite NRC Emergency Action Level (EAL):

Communications Plant Radio X

1. Loss of ALL Table C1 Onsite communications Plant Page X capability affecting the ability to perform routine All telephone Lines X X X operations.

(Commercial and microwave) OR ENS X X 2. Loss of ALL Table C1 Offsite communication HPN X X capability affecting the ability to perform offsite notifications.

Satellite Phones X X OR

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

CA5 Inability to maintain plant in cold 4 5 CU5 UNPLANNED rise in RCS temperature. 4 5 Table C2 RCS Heat-up Duration Thresholds shutdown RCS Containment Closure Heat-up Emergency Action Levels (EAL):

Emergency Action Levels (EAL):

Status Status Duration Intact Not Applicable 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 has been exceeded, or will likely be exceeded. has been exceeded, or will likely be exceeded.

Established 20 minutes*

Heat Sink Not Intact A momentary UNPLANNED excursion above the A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature Technical Specification cold shutdown temperature limit when heat removal function is available does not limit when heat removal function is available does not Not Established 0 minutes warrant classification. warrant classification.

1. UNPLANNED rise in RCS temperature > 200ºF for 1. UNPLANNED rise in RCS temperature > 200ºF.

If an RCS heat removal system is in operation within > Table C2 duration. OR this time frame and RCS temperature is being OR 2. Loss of the following for > 15 minutes.

reduced, then EAL Threshold #1 is not applicable.

2. UNPLANNED RCS pressure rise > 10 psig as a

ALL RCS temperature indications result of temperature rise.

AND

ALL RPV level indications Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-17 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Cold Shutdown / Refueling System Malfunctions CG6 Loss of RPV inventory affecting 4 5 CS6 Loss of RPV inventory affecting core 4 5 CA6 Loss of RPV inventory 4 5 CU6 UNPLANNED loss of RPV inventory 4 5 fuel clad integrity with containment challenged. decay heat removal capabilities. for 15 minutes or longer.

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

Emergency 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 has promptly upon determining that the applicable time 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.

been exceeded, or will likely be exceeded.

1 a. RPV water Level < -161 inches (TAF) for > 30 minutes. 1. With CONTAINMENT CLOSURE not established, 1. Loss of RPV inventory as indicated by level

< - 83 inches (Level 2). 1. UNPLANNED loss of reactor coolant results in the inability AND RPV water level < -147inches (Level 1) to restore and maintain RPV water level to above the

b. Any Containment Challenge Indication (Table C4) OR procedurally established lower limit for > 15 minutes.

OR OR

2. With CONTAINMENT CLOSURE established, RPV 2. a. RPV water level cannot be monitored for OR

2. a. RPV water level cannot be monitored for > 30 minutes.

water level < - 161 inches (TAF) > 15 minutes.

AND 2. a. RPV water level cannot be monitored

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

Table C3 indications of a sufficient magnitude 3. a. RPV water level cannot be monitored for b. Loss of RPV inventory per Table C3 to indicate core uncovery. indications b. Loss of RPV inventory per Table C3 indications.

> 30 minutes OR RCS Leakage / Inventory AND

Refuel floor Rad monitor 0D21-K604A

>3000 mR/hr b.Core uncovery is indicated by ANY of the AND following:

Table C3 indications of a sufficient

3. ANY Containment Challenge Indication (Table C4) magnitude to indicate core uncovery.

OR

Refuel Floor Rad monitor 0D21-K604A

>3000 mR/hr.

Table C3 Indications of RCS Leakage

UNPLANNED floor or equipment sump level rise*

UNPLANNED Suppression Pool level rise*

UNPLANNED vessel make up rate rise*

Observation of leakage or inventory loss

Rise in level is attributed to a loss of RPV inventory Table C4 Containment Challenge Indications

Primary Containment Hydrogen Concentration >6%

and Oxygen >5%

UNPLANNED rise in containment pressure

CONTAINMENT CLOSURE not established*

ANY Secondary Containment radiation monitor

> LGA-002 Maximum Safe operating level.

if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-18 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HG1 HOSTILE ACTION resulting in loss 1 2 3 4 5 D HS1 HOSTILE ACTION within the 1 2 3 4 5 D HA1 HOSTILE ACTION within the 1 2 3 4 5 D HU1 Confirmed SECURITY CONDITION 1 2 3 4 5 D of physical control of the facility PROTECTED AREA OWNER CONTROLLED AREA or airborne attack or threat.

threat within 30 minutes.

Emergency Action Level (EAL):

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

Hostile Action

1. A notification from the Security Force that a HOSTILE 1. Notification of a credible security threat directed at A notification from the Security Force that a HOSTILE 1. A validated notification from NRC of an aircraft ACTION is occurring or has occurred within the the site as determined per SY-AA-101-132, Security ACTION is occurring or has occurred within the attack threat < 30 minutes from the site.

PROTECTED AREA. Assessment and Response to Unusual Activities.

PROTECTED AREA.

AND OR OR

2. a. ANY Table H1 safety function cannot be 2. Notification by the Security Force that a HOSTILE 2. A validated notification from the NRC providing controlled or maintained. ACTION is occurring or has occurred within the information of an aircraft threat.

OWNER CONTROLED AREA. OR OR

3. Notification by the Security Force of a SECURITY

b. Damage to spent fuel has occurred or is CONDITION that does not involve a HOSTILE IMMINENT ACTION.

HS2 Inability to control a key safety 1 2 3 4 5 D HA2 Control Room evacuation resulting 1 2 3 4 5 D function from outside the Control Room in transfer of plant control to alternate locations Table H1 Safety Functions Emergency Action Level (EAL): Emergency Action Level (EAL):

Reactivity Control Note: The Emergency Director should declare the event An event has resulted in plant control being transferred from (ability to shutdown the reactor and keep it shutdown) promptly upon determining that the applicable time has the Control Room to alternate locations per:

been exceeded, or will likely be exceeded.

LOA-RX-101(201), Unit 1(2) Control room

RPV Water Level (ability to cool the core)

1. An event has resulted in plant control being transferred Evacuation Abnormal Transfer of Plant Control from the Control Room to alternate locations per: OR

RCS Heat Removal (ability to maintain a heatsink)

LOA-RX-101(201), Unit 1(2) Control room Evacuation

LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Abnormal Fire in the Control Room OR AEER.

OR

LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Fire in the Control Room OR AEER.

AND

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 October 2017 COLD MATRIX LS 2-19 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HU3 FIRE potentially degrading the level 1 2 3 4 5 D of safety of the plant.

Emergency Action Level (EAL):

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

Escalation of the emergency classification level would be via IC CA2 or MA5 Table H2 Vital Areas

Reactor Building 1. A FIRE in ANY Table H2 area is not extinguished in (when inerted the Drywell is exempt) < 15-minutes of ANY of the following FIRE detection

Control Room indications:

Fire

Auxiliary Building

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

Receipt of multiple (more than 1) fire alarms or

Unit and Shared Emergency Diesel indications Generator Rooms

Field verification of a single fire alarm

Switchgear and Battery Rooms OR

Remote Shutdown Rooms 2. a. Receipt of a single fire alarm in ANY Table H2 area

CSCS Pump Rooms (i.e., no other indications of a FIRE).

LSH (for 0E12-F300 access only) 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 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-20 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HU4 Seismic event greater than OBE levels 1 2 3 4 5 D Emergency Action Level (EAL):

Note: Escalation of the emergency classification level would be via IC CA2 or MA5 For emergency classification if EAL # 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in < 15 mins of the event.

1. Seismic event > Operating Basis Earthquake (OBE) as indicated by:

Alarm on Panel 1PM10J B503 SEISMIC OBE/SSE Earthquake LEVEL EXCEEDED OR

OBE LED lit on 0PA11J 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 mins of the event:

The earthquake resulted in Modified Mercalli Intensity (MMI) > VI and occurred < 3.5 miles of the plant.

The earthquake was magnitude > 6.0

The earthquake was magnitude > 5.0 and occurred

< 125 miles of the plant.

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-21 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety Table H3 HA5 Gaseous release impeding access to 3 4 Areas with Entry Related Mode Applicability equipment necessary for normal plant operations, Entry Related Mode cooldown or shutdown Area Applicability Emergency Action Level (EAL):

Reactor Building 673 SE Corner Room Note: If the equipment in the listed room or area was already inoperable, or out of service, before the Toxic Gas 694 NW Corner Room SE Corner Room event occurred, then no emergency classification Raceway is warranted.

Mode 3 and 4 740 NE Valve Room 1. Release of a toxic, corrosive, asphyxiant or NE flammable gas in ANY Table H3 area.

SW 761 NE AND SW 2. Entry into the room or area is prohibited or impeded Auxiliary Building 674 CSCS Rooms Mode 3 and 4 731 AEER 736 DG Penthouse HU6 Hazardous Event 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 Hazardous Event

1. Tornado strike within the PROTECTED AREA.

OR

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

OR

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

OR

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

Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-22 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Hazards and Other conditions Affecting Plant Safety HG7 Other conditions exist which in the 1 2 3 4 5 D HS7 Other conditions exist which in the 1 2 3 4 5 D HA7 Other conditions exist which in the 1 2 3 4 5 D HU7 Other conditions exist which in the 1 2 3 4 5 D judgment of the Emergency Director warrant judgment of the Emergency Director warrant judgment of the Emergency Director warrant judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY. declaration of a SITE AREA EMERGENCY. declaration of an ALERT. declaration of an UNUSUAL 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 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 Emergency Director, indicate that events are in progress or 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 have occurred which involve an actual or potential 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 substantial degradation of the level of safety of the plant or a 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, security event that involves probable life threatening risk to 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 site personnel or damage to site equipment because of 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 HOSTILE ACTION. Any releases are expected to be limited degradation of safety systems occurs.

Guideline exposure levels offsite for more than the immediate equipment needed for the protection of the public. Any to small fractions of the EPA Protective Action Guideline site area. releases are not expected to result in exposure levels which exposure levels.

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

Emergency Director Judgment Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-23 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex COLD MATRIX COLD MATRIX Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ISFSI Malfunction E-HU1 Damage to a loaded cask 1 2 3 4 5 D CONFINEMENT BOUNDARY.

Emergency Action Level (EAL):

Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by exceeding ANY Table E1 on-contact radiation reading.

Table E1 On-contact radiation readings ISFSI Cask Type Location of on-contact radiation reading Top of the Side of the spent fuel spent fuel cask, excluding inlet HI-STORM cask and outlet ducts (mrem/hr) (mrem/hr)

Labeled as

> 40 > 220 XXX-A3 Labeled as

> 60 > 600 XXX-A8 Modes: 1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled October 2017 COLD MATRIX LS 2-24 COLD MATRIX EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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

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

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

1. The sum of readings on the Vent Stack and SBGT WRGMs > 8.26E+09 uCi/sec for

> 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR

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

a. > 1000 mRem TEDE OR

b. > 5000 mRem CDE Thyroid OR

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

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

OR

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

October 2017 LS 2-25 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG1 (cont)

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

The TEDE dose is set at the EPA PAG of 1000 mRem while the 5000 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Basis Reference(s):

1. NEI 99-01 Rev 6, AG1

2. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems

3. ODCM Section 12.4 Gaseous Effluents and Total Dose

4. EP-EAL-0605, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values LaSalle Station

5. EP-AA-112-500, Emergency Environmental Monitoring October 2017 LS 2-26 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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

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

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

1. The sum of readings on the Vent Stack and SBGT WRGMs > 8.26 E+08 uCi/sec for

> 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR

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

a. > 100 mRem TEDE OR

b. > 500 mRem CDE Thyroid OR

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

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

OR

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

October 2017 LS 2-27 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS1 (cont)

Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs).

It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

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

The TEDE dose is set at 10% of the EPA PAG of 1000 mRem while the 500 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, AS1

2. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems

3. ODCM Section 12.4 Gaseous Effluents and Total Dose

4. EP-EAL-0605, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values LaSalle Station

5. EP-AA-112-500, Emergency Environmental Monitoring October 2017 LS 2-28 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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

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

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

1. The sum of readings on the Vent Stack and SBGT WRGMs > 8.26 E+07 uCi/sec for

> 15 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR

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

a. > 10 mRem TEDE OR

b. > 50 mRem CDE Thyroid OR

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 October 2017 LS 2-29 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 (cont)

Emergency Action Level (EAL) (cont):

OR

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

a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for > 60 minutes.

OR

b. Analyses of field survey s a m p l e s i n d i c a t e > 50 mRem CDE Thyroid for 60 minutes of inhalation.

Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

The TEDE dose is set at 1% of the EPA PAG of 1000 mRem while the 50 mRem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

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

October 2017 LS 2-30 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA1 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, AA1

2. ODCM Section 12.3 Liquid Effluents

3. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems

4. ODCM Section 12.4 Gaseous Effluents and Total Dose

5. EP-EAL-0605, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values LaSalle Station

6. LCP-140-7, Analysis Of Radwaste Discharge Tanks 1(2)WF05T and Determination Of Discharge Flowrate And Liquid Radwaste Effluent Monitor

Response

7. Structural Drawing S-01A Composite Site Plan LaSalle Station Units 1 & 2

8. LAP-1800-4, Chemistry Department Improved Technical Specifications, Technical Requirements Manual, TRM Appendixes, Offsite Dose Calculation Manual (ODCM) LaSalle Annex Check Lists

9. LYP-1200-2, Instantaneous Airborne Releases - 10 CFR 20 Design Objectives

10. EP-EAL-0613, LaSalle Criteria for Choosing Radiological Liquid Effluent EAL Threshold Values October 2017 LS 2-31 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 Initiating Condition:

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

Operating Mode Applicability:

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

Notes:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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

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

1. Reading on ANY of the following effluent monitors > 2 times alarm setpoint established by a current radioactive release discharge permit for 60 minutes.

Radwaste Effluent Monitor LCRM 0D18-K606 OR

Discharge Permit specified monitor OR

2. The sum of readings on the Vent Stack and SBGT WRGMs > 9.42 E+05 uCi/sec for > 60 minutes (as determined by Control Room Panels or PPDS - Total Noble Gas Release Rate).

OR

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

Basis:

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

October 2017 LS 2-32 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 (cont)

Basis (cont):

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

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

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

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

The effluent monitors listed are those normally used for planned discharges. If a discharge is performed using a different flowpath or effluent monitor other than those listed (e.g., a portable or temporary effluent monitor), then the declaration criteria will be based on the monitor specified in the Discharge Permit.

EAL #2 Basis This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous effluent pathways.

EAL #3 Basis This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g.,

spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC RA1.

October 2017 LS 2-33 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU1 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, AU1

2. UFSAR Section 11.5, Process and Effluent Radiological Monitoring and Sampling Systems

3. ODCM Section 12.4 Gaseous Effluents and Total Dose

4. ODCM Section 12.3, Liquid Effluents

5. LCP-140-7, Analysis of Radwaste Discharge Tanks 1(2)WF05T and Determination of Discharge Flowrate and Liquid Radwaste Effluent Monitor

Response

6. EP-EAL-0605, Criteria for Choosing Radiological Gaseous Effluent EAL Threshold Values LaSalle Station

7. LAP-1800-4, Chemistry Department Improved Technical Specifications, Technical Requirements Manual, TRM Appendixes, Offsite Dose Calculation Manual (ODCM) LaSalle Annex Check Lists

8. LYP-1200-2, Instantaneous Airborne Releases - 10 CFR 20 Design Objectives October 2017 LS 2-34 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RG2 Initiating Condition:

Spent fuel pool level cannot be restored to at least 820.42 ft. as indicated on 1(2) LI-FC165 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 General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

Spent fuel pool level cannot be restored to at least 820.42 ft. as indicated on 1(2) LI-FC165 for 60 minutes or longer.

Basis:

This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, AG2

2. EP-EAL-1000, Criteria for Choosing Spent Fuel Pool Level 3 EAL Threshold Values for LaSalle Station October 2017 LS 2-35 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RS2 Initiating Condition:

Spent fuel pool level at 820.42 ft. as indicated on 1(2) LI-FC165.

Operating Mode Applicability:

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

Lowering of spent fuel pool level to 820.42 ft. as indicated on 1(2) LI-FC165.

Basis:

This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

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

Escalation of the emergency classification level would be via IC RG1 or RG2.

Basis Reference(s):

1. NEI 99-01 Rev 6, AS2

2. EP-EAL-1000, Criteria for Choosing Spent Fuel Pool Level 3 EAL Threshold Values for LaSalle Station October 2017 LS 2-36 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 Initiating Condition:

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

Operating Mode Applicability:

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

1. Uncovery of irradiated fuel in the REFUELING PATHWAY.

OR

2. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ARM 0D21-K604A Radiation Monitor reading >1000 mRem/hr.

OR

3. Lowering of spent fuel pool level to 830.00 ft. as indicated on 1(2) LI-FC165.

Basis:

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

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

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

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HU1.

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.

October 2017 LS 2-37 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA2 (cont)

Basis (cont):

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.

Basis Reference(s):

1. NEI 99-01 Rev 6, AA2

2. Technical Specification 3.9.6 Reactor Pressure Vessel (RPV) Water Level Irradiated Fuel

3. Technical Specification 3.7.8 Spent Fuel Storage Pool Water Level

4. LOA-FH-001 Irradiated Fuel Assembly Damage

5. LOR-1(2)H13-P601-B108, Refuel Floor Area High Range/Low Range Radiation -

High

6. LOR-1(2)H13-P601-E205/F205, Fuel Pool Radiation High High

7. LOP-SF-06 Filling the Reactor, Reactor Well and Dryer/Separator Pit Through Feedwater with Suppression Pool Cleanup

8. LOA-FC-101(201), Unit 1(2) Fuel Pool Cooling System Abnormal

9. LOA-AR-101(201), Area Radiation Monitoring System Abnormal

10. LaSalle Design Consideration Summary EC 394010 Rev 1, Spent Fuel Pool Level Instrumentation - Fukushima October 2017 LS 2-38 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 Initiating Condition:

UNPLANNED loss of water level above irradiated fuel.

Operating Mode Applicability:

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

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

Refueling Cavity water level < 340 in. on shutdown range.

OR

Spent Fuel Pool water level < 21 ft. 4 in.

OR

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

AND

b. UNPLANNED Area Radiation Monitor reading rise on ARM 0D21-K604A radiation monitor.

Basis:

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

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

This IC addresses a loss in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

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

October 2017 LS 2-39 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU2 (cont)

Basis (cont):

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

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

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

Basis Reference(s):

1. NEI 99-01 Rev 6, AU2

2. LRP-5800-3 Radiation Monitoring Alarm/Trip Setpoint Determination

3. RP-AA-203, Exposure Control and Authorization

4. Technical Specification 3.7.8, Spent Fuel Storage Pool Water Level

5. Technical Specification 3.9.6, Reactor Pressure Vessel (RPV) Water Level Irradiated Fuel

6. Technical Specification 3.9.7, Reactor Pressure Vessel (RPV) Water Level New Fuel or Control Rods

7. LOP-SF-06 Filling the Reactor, Reactor Well and Dryer/Separator Pit Through Feedwater with Suppression Pool Cleanup

8. LOA-FC-101(201), Unit 1(2) Fuel Pool Cooling System Abnormal October 2017 LS 2-40 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA3 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

If the equipment in the room or area listed in Table R3 was already inoperable, or out of service, before the event occurred, then no emergency classification is warranted.

1. Dose rate > 15 mR/hr in ANY of the following Table R2 areas:

Table R2 Areas Requiring Continuous Occupancy

Main Control Room (1(2)D18-K751A-D)

Central Alarm Station - (by survey)

OR

2. UNPLANNED event results in radiation levels that prohibit or significantly impede access to ANY of the following Table R3 plant rooms or areas:

Table R3 Areas with Entry Related Mode Applicability Entry Related Mode Area Applicability Reactor Building 673 SE Corner Room 694 NW Corner Room SE Corner Room Raceway Mode 3 and 4 740 NE Valve Room NE SW 761 NE SW Auxiliary Building 674 CSCS Rooms Mode 3 and 4 731 AEER 736 DG Penthouse October 2017 LS 2-41 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA3 (cont)

Basis:

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

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

Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR, therefore Table R3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where if this action is not completed the plant would not be able to attain and maintain cold shutdown.

This Table does not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

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

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

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

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

October 2017 LS 2-42 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RA3 (cont)

Basis (cont):

The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, AA3

2. LRP-5800-3, Radiation Monitoring Alarm/Trip Setpoint Determination

3. LIS-AR-105 (205)A-D, Main Control Room Radiation Monitor Channel A Calibration

4. UFSAR Section 3.8

5. UFSAR Section 12.3.2.5 October 2017 LS 2-43 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENTS RU3 Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability:

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

1. Offgas system radiation monitor HI-HI alarm.

OR

2. Specific coolant activity > 4.0 uCI/gm Dose Equivalent I-131.

Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Conditions that cause the specified monitor to alarm that are not related to fuel clad degradation should not result in the declaration of an Unusual Event.

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

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

Basis Reference(s):

1. NEI 99-01 Rev 6, SU3

2. Technical Specifications 3.4.8

3. LOR 1(2) N62-P600-B501, Off Gas Pre-Treatment Radiation Monitor High-High Radiation

4. LRP-5820-34, Off-Gas Pre and Post Treatment Monitor Alarm and Trip Setpoints

5. LIS-OG-102 (202), Steam Jet Air Ejector Off Gas (Pretreatment) Radiation Monitor Calibration

6. LOA-AR-101 (201), Area Radiation Monitoring System Abnormal

7. LAP-1800-4, Chemistry Department Improved Technical Specifications, Technical Requirements Manual, TRM Appendixes, Offsite Dose Calculation Manual (ODCM) LaSalle Annex Check Lists October 2017 LS 2-44 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-45 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-46 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-47 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 Dose Equivalent I-131.

Basis:

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

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with RCS Activity.

Basis Reference(s):

1. NEI 99-01 Rev 6, Table 9-F-2 October 2017 LS 2-48 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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. Plant conditions indicate Primary Containment flooding is required.

POTENTIAL LOSS

2. RPV water level cannot be restored and maintained > -161 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 SAMG program.

Loss Threshold #1 Basis The Loss threshold represents the EOP requirement for primary containment flooding.

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

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

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

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water level cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate October 2017 LS 2-49 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FC2 (cont)

Basis (cont):

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

2. LGA-001, RPV Control

3. LGA-010, Failure to Scram

4. LGA-005, RPV Flooding

5. LPGP-CALC-02, EOP & SAMG Calculation Control -- Setpoints and Calculation Results October 2017 LS 2-50 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 > 1.90 E+02 R/hr (190 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 Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier RC5 Loss Threshold since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

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

Basis Reference(s):

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

2. Core Damage Assessment Methodology October 2017 LS 2-51 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-52 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 > -161 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 SAMG 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.

October 2017 LS 2-53 EP-AA-1005 Addendum 3 (Revision 3)

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

2. LPGP-CALC-02, EOP & SAMG Calculation Control -- Setpoints and Calculation Results

3. LGA-001, RPV Control October 2017 LS 2-54 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC3 Initiating Condition:

Primary Containment Pressure Operating Mode Applicability:

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

LOSS

1. Drywell pressure >1.77 psig.

AND

2. Drywell pressure rise is due to RCS leakage Basis:

The > 1.77 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.

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

Basis Reference(s):

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

2. UFSAR Table 3.3.5.1-1

3. Technical Specifications Table 3.3.5.1-1

4. LGA-001, RPV Control

5. LGA-003, Primary Containment Control October 2017 LS 2-55 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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), RCIC, Feedwater, or RWCU line break.

OR

2. Emergency RPV Depressurization is required.

POTENTIAL LOSS

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

a. Secondary Containment area temperature > LGA-002 Maximum Normal operating levels.

OR

b. Secondary Containment area radiation level > LGA-002 Maximum Normal operating level.

Basis:

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

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

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

Reports describing magnitude of steam or water release.

Use of system high flow alarms / indications, if available,

Significant changes in makeup requirements,

Abnormal reactor water level changes in response to the event.

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

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

October 2017 LS 2-56 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC4 (cont)

Basis (cont):

Loss Threshold #2 Basis Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the 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 Operating values in areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment.

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

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

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

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

1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Basis Reference(s):

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

2. UFSAR Section 5.4.9

3. Technical Specifications 3.4.5 RCS Operational LEAKAGE

4. UFSAR Section 5.2.5

5. LOP-NB-03, Troubleshooting Drywell Leakage

6. LGA-002, Secondary Containment Control October 2017 LS 2-57 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION RC5 Initiating Condition:

Primary Containment Radiation Operating Mode Applicability:

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

LOSS

1. Drywell radiation monitor reading >1.00 E+02 R/hr (> 100R/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. NEI 99-01 Rev 6, Table 9-F-2

2. Calc. EP-EAL-0611 October 2017 LS 2-58 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-59 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 Plant conditions indicate Primary Containment flooding is required.

Basis:

The Potential Loss threshold is identical to the Fuel Clad Barrier FC2 Loss threshold RPV Water Level. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAMGs specify the conditions that require primary containment flooding. When primary containment flooding is required, the EPGs are exited and SAMGs are entered. Entry into SAMGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

Basis Reference(s):

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

2. LGA-001, RPV Control

3. LGA-010, Failure to Scram

4. LGA-005, RPV Flooding October 2017 LS 2-60 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT3 Initiating Condition:

Primary Containment Conditions Operating Mode Applicability:

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

LOSS

1. UNPLANNED rapid drop in Drywell pressure following Drywell pressure rise.

OR

2. Drywell pressure response not consistent with LOCA conditions.

POTENTIAL LOSS

3. Drywell pressure > 45 psig and rising.

OR

4. a. Drywell or suppression chamber hydrogen concentration > 6%.

AND

b. Drywell or suppression chamber oxygen concentration > 5%.

OR

5. Heat Capacity Limit (LGA-003, Fig.H) 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.

October 2017 LS 2-61 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR Suppression chamber pressure above Potential Loss Threshold #3 Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

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

Basis Reference(s):

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

2. UFSAR 15.6.5

3. UFSAR Section 6.2

4. LGA-011, Hydrogen Control

5. LGA-003, Primary Containment Control

6. LaSalle PSTG Section 5B, Hydrogen Control October 2017 LS 2-62 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 > 4.35 E+02 R/hr ( > 435 R/hr).

Basis:

There is no Loss threshold associated with Primary Containment Radiation.

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

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

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

Basis Reference(s):

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

2. Core Damage Assessment Methodology October 2017 LS 2-63 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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/purging per EOPs or SAMGs due to accident conditions.

OR

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

a. Secondary Containment area temperature > LGA-002, Maximum Safe operating levels.

OR

b. Secondary Containment area radiation level > LGA-002, Maximum Safe operating levels.

Basis:

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

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

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

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

Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main Steamline, RCIC steamline breaks, unisolable RWCU system breaks, and unisolable containment atmosphere vent paths.

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

October 2017 LS 2-64 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

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

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

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

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

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the Drywell high pressure scram setpoint) does not meet the threshold condition.

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

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

In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Reactor Building since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the Reactor Building, an unexpected rise in Feedwater flowrate, or unexpected October 2017 LS 2-65 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CT6 (cont)

Basis (cont):

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

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

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

Basis Reference(s):

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

2. LGA-002, Secondary Containment Control October 2017 LS 2-66 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-67 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. Loss of ALL offsite AC power to unit ECCS busses(excluding Division 3).

AND

2. Failure of DG 0 and DG 1A(2A) emergency diesel generators to supply power to unit ECCS busses.

AND

3. EITHER of the following:

a. Restoration of at least one unit ECCS bus (excluding Division 3) in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely.

OR

b. RPV water level cannot be restored and maintained > -150 inches Basis:

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

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

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

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.

October 2017 LS 2-68 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont)

Basis (cont):

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Basis Reference(s):

1. NEI 99-01 Rev 6, SG1

2. UFSAR 8.1

3. LOA-AP-101(201), Unit 1(2) AC Power System Abnormal

4. UFSAR 15.9

5. LPGP-CALC-02, EOP & SAMG Calculation Control -- Setpoints and Calculation Results

6. LGA-001, RPV Control

7. LGA-010, Failure to Scram October 2017 LS 2-69 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. Loss of ALL offsite AC Power to unit ECCS busses(excluding Division 3).

AND

2. Failure of DG 0 and DG 1A(2A) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one unit ECCS bus (excluding Division 3) in

< 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

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

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, SS1

2. UFSAR 8.1.

3. LOA-AP-101(201), Unit 1(2) AC Power System Abnormal October 2017 LS 2-70 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. AC power capability to unit ECCS busses (excluding Division 3) reduced to only one of the following power sources for > 15 minutes.

System Auxiliary Transformer 142(242)

Unit Auxiliary Transformer 141(241)

Unit Emergency Diesel Generator DG 1A(2A)

Shared Emergency Diesel Generator DG 0

Other SAT via crosstie breakers AND

2. ANY additional single power source failure (excluding Division 3) 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.

A loss of all offsite power (to both units) with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

October 2017 LS 2-71 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 (cont)

Basis (cont):

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from the unit main generator via the UAT.

A loss of the UAT and SAT and a loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source via the other unit.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Escalation of the emergency classification level would be via IC MS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA1

2. UFSAR 8.1

3. LOA-AP-101(201) Unit 1(2) AC Power System Abnormal

4. UFSAR 15.9 October 2017 LS 2-72 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU1 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

Loss of ALL offsite AC power capability to unit ECCS busses (excluding Division 3) for

> 15 minutes.

Basis:

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

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

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

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC MA1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU1

2. UFSAR 8.1

3. LOA-AP-101 (201), Unit 1(2) AC Power System Abnormal October 2017 LS 2-73 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. Loss of ALL offsite AC power to unit ECCS busses (excluding Division 3).

AND

2. Failure of DG 0 and DG 1A(2A) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Voltage is < 108 VDC on unit 125 VDC battery busses 111Y(211Y) and 112Y(212Y).

AND

4. ALL AC and Vital DC power sources (excluding Division 3) 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.

October 2017 LS 2-74 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG2 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, SG8

2. UFSAR 8.3.2

3. UFSAR 8.1

4. LOA-DC-101(201) Unit 1(2) DC Power System Failure

5. LOA-AP-101(201), Unit 1(2) AC Power System Abnormal October 2017 LS 2-75 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

Voltage is < 108 VDC on 125 VDC battery busses 111Y(211Y) and 112Y(212Y) for > 15 minutes.

Basis:

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

This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, SS8

2. UFSAR 8.3.2.1.1

3. Technical Specifications B3.8.4

4. UFSAR 8.3.2

5. LOA-DC-101(201) Unit 1(2) DC Power System Failure October 2017 LS 2-76 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 > 3%.

AND

2. ALL manual / ARI actions to shutdown the reactor have been unsuccessful as indicated by Reactor Power > 3%.

AND

3. EITHER of the following conditions exist:

RPV water level cannot be restored and maintained > -150 inches on WR (-183 inches (MSCRWL) on FZ if WR not available)

OR

Heat Capacity Limit (LGA-003, Fig. H) exceeded.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator manual actions, both inside and outside the Control Room including driving in control rods and boron injection, are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

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

Escalation of the emergency classification level would be via IC RG1 or FG1.

October 2017 LS 2-77 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, SS5

2. LGA-010, Failure to Scam

3. LPGP-CALC-02, EOP & SAMG Calculation Control - Setpoints and Calculation Results

4. LGA-003, Primary Containment Control October 2017 LS 2-78 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 Initiating Condition:

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

Operating Mode Applicability:

1, 2 Emergency Action Level (EAL):

Note:

A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

1. Automatic or manual scram did not shutdown the reactor as indicated by Reactor Power > 3%.

AND

2. Manual / ARI actions taken at the Reactor Console are not successful in shutting down the reactor as indicated by Reactor Power > 3%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles.

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.

October 2017 LS 2-79 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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. NEI 99-01 Rev 6, SA5

2. LGA-001, RPV Control

3. LGA-010, Failure to Scram

4. LGA-003, Primary Containment Control October 2017 LS 2-80 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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, and does not include manually driving in control rods or implementation of boron injection strategies.

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

> 3%.

AND

b. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor as indicated by Reactor Power < 3%.

OR

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

> 3%.

AND

b. EITHER of the following:

1. Subsequent manual / ARI action taken at the Reactor Console is successful in shutting down the reactor as indicated by Reactor Power

< 3%.

OR

2. Subsequent automatic scram / ARI is successful in shutting down the reactor as indicated by Reactor Power < 3%.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

October 2017 LS 2-81 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU3 (cont)

Basis (cont):

EAL #1 Basis Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram/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 plants decay heat removal systems.

EAL #2 Basis If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram/ARI using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram/ARI is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles.

Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MA3. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC MA3 or FA1, an Unusual Event declaration is appropriate for this event.

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.

October 2017 LS 2-82 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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. NEI 99-01 Rev 6, SU5

2. LGA-001, RPV Control

3. LGA-010, Failure to Scram

4. Technical Specifications Table 3.3.1.1-1 October 2017 LS 2-83 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. UNPLANNED event results in the inability to monitor ANY Table M1 parameters from within the Control Room for >15 minutes.

Table M1 Control Room Parameters

Reactor Power

RPV Water Level

RPV Pressure

Primary Containment Pressure

Suppression Pool Level

Suppression Pool Temperature AND

2. ANY Table M2 transient in progress.

Table M2 Significant Transients

Turbine Trip

Reactor Scram

ECCS Activation

Recirc. Runback > 25% Reactor Power Change

Thermal Power oscillations > 10% Reactor Power Change October 2017 LS 2-84 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine any of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for any of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RS1.

Basis Reference(s):

1. NEI 99-01 Rev 6, SA2 October 2017 LS 2-85 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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

Table M1 Control Room Parameters

Reactor Power

RPV Water Level

RPV Pressure

Primary Containment Pressure

Suppression Pool Level

Suppression Pool Temperature Basis:

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

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

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an inability to monitor means that values for any of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

October 2017 LS 2-86 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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. NEI 99-01 Rev 6, SU2 October 2017 LS 2-87 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 Initiating Condition:

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

Operating Mode Applicability:

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

Note:

If it is determined that the conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.

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

Seismic event (earthquake)

Internal or external flooding event

High winds or tornado strike

FIRE

EXPLOSION

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

2. EITHER of the following:

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

OR

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

Basis:

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

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

October 2017 LS 2-88 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA5 (cont)

Basis (cont):

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

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, required for the current operating mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. Manual or automatic electrical isolation of safety equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.

EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

EAL #2.b addresses damage to a SAFETY SYSTEM component that is required to be operable by Technical Specifications for the current operating mode, and is not in operation or readily apparent through indications alone, as well as damage to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC FS1 or RS1.

If the EAL conditions of MA5 are not met then assess the event via HU3, HU4, or HU6.

Basis Reference(s):

1. NEI 99-01, Rev 6 SA9 October 2017 LS 2-89 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. RCS unidentified or pressure boundary leakage in the Drywell > 10 gpm for

> 15 minutes.

OR

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

OR

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

Basis:

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

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

This IC addresses RCS leakage which may be a precursor to a more significant event.

In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

EAL #1 and EAL #2 Basis These EALs are focused on a loss of mass from the RCS due to unidentified leakage",

"pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications).

EAL #3 Basis This EAL addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system.

These three EALs thus apply to leakage into the containment, a secondary-side system or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming October 2017 LS 2-90 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 (cont)

Basis (cont):

calculations to determine (e.g., a mass balance calculation). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

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

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

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

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

Basis Reference(s):

1. NEI 99-01 Rev 6, SU4

2. Technical Specifications 3.4. 5

3. UFSAR 5.2.5

4. LOP-NB-03, Troubleshooting Drywell Leakage

6. LGA-001, RPV Control October 2017 LS 2-91 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 Initiating Condition:

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

Operating Mode Applicability:

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

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

OR

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

OR

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

Table M3 Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:

This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

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

October 2017 LS 2-92 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 (cont)

Basis (cont):

EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are listed in procedure EP-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) Form.

EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Basis Reference(s):

1. NEI 99-01 Rev 6, SU6

2. EP-MW-124-1001 Facilities Inventories and Equipment Tests

3. LAP-100-37, Station Communications

4. LOP-CQ-02, Intercom/Loud Speaker System

5. LOP-CQ-03, Sound Powered Telephone System

6. LOP-CQ-04, Intra-Plant Radio System

7. OP-AA-104-101, Communications October 2017 LS 2-93 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 Initiating Condition:

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

Operating Mode Applicability:

4, 5, D Emergency Action Level (EAL):

Note:

The Emergency Director should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. Loss of ALL offsite AC power to unit ECCS busses (excluding Division 3).

AND

2. Failure of DG 0 and DG 1A(2A) emergency diesel generators to supply power to unit ECCS busses.

AND

3. Failure to restore power to at least one unit ECCS bus (excluding Division 3) in

< 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC CS6 or RS1.

October 2017 LS 2-94 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA1 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, CA2

2. UFSAR 8.3

3. LOA-AP-101 (201), Unit 1(2) AC Power System Abnormal October 2017 LS 2-95 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. AC power capability to unit ECCS busses (excluding Division 3) reduced to only one of the following power sources for > 15 minutes.

System Auxiliary Transformer 142(242)

Unit Auxiliary Transformer 141(241)

Unit Emergency Diesel Generator DG 1A(2A)

Shared Emergency Diesel Generator DG 0

Other SAT via crosstie breakers AND

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

Basis:

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

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

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

October 2017 LS 2-96 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 (to both units) with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from the unit main generator via the UAT.

A loss of the UAT and SAT and a loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source via the other unit.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA1.

Basis Reference(s):

1. NEI 99-01 Rev 6 CU2

2. UFSAR 8.1

3. LOA-AP-101 (201) Unit 1(2) AC Power System Abnormal

4. UFSAR 15.9 October 2017 LS 2-97 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 Initiating Condition:

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

Operating Mode Applicability:

4, 5 Emergency Action Level (EAL):

Note:

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

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

Seismic event (earthquake)

Internal or external flooding event

High winds or tornado strike

FIRE

EXPLOSION

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

2. EITHER of the following:

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

OR

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

Basis:

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

October 2017 LS 2-98 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA2 (cont)

Basis (cont):

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

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

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, required for the current operating mode, "required", i.e. required to be operable by Technical Specifications for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. Manual or automatic electrical isolation of safety equipment due to flooding, in and of itself, does not constitute degraded performance and is classified under HU6.

EAL #2.a addresses damage to a SAFETY SYSTEM train that is required to be operable by Technical Specifications for the current operating mode, and is in operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

EAL #2.b addresses damage to a SAFETY SYSTEM component that is required to be operable by Technical Specifications for the current operating mode, and is not in operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC CS6 or RS1.

If the EAL conditions of CA2 are not met then assess the event via HU3, HU4, or HU6.

Basis Reference(s):

1. NEI 99-01 Rev 6, CA6 October 2017 LS 2-99 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

Voltage is < 108 VDC on required unit 125 VDC battery busses 111Y(211Y) and 112Y(212Y) for > 15 minutes.

Basis:

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

This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, required means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.

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

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Depending upon the event, escalation of the emergency classification level would be via IC CA6 or CA5, or an IC in Recognition Category R.

Basis Reference(s):

1. NEI 99-01 Rev 6, CU4

2. UFSAR 8.3.2

3. LOA-DC-101(201), Unit 1(2) DC Power System Failure October 2017 LS 2-100 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

4, 5, D Emergency Action Level (EAL):

1. Loss of ALL Table C1 Onsite communications capability affecting the ability to perform routine operations.

OR

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

OR

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

Table C1 Communications Capability System Onsite Offsite NRC Plant Radio X Plant Page X All telephone Lines (Commercial and microwave) X X X ENS X X HPN X X Satellite Phones X X Basis:

This IC addresses a significant loss of on-site, offsite, or NRC communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

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

October 2017 LS 2-101 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU4 (cont)

Basis (cont):

EAL #2 Basis Addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are listed in procedure EP-MW-114-100-F-01, Nuclear Accident Reporting System (NARS) Form.

EAL #3 Basis Addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Basis Reference(s):

1. NEI 99-01 Rev 6, CU5

2. EP-MW-124-1001 Facilities Inventories and Equipment Tests

3. LAP-100-37, Station Communications

4. LOP-CQ-02, Intercom/Loud Speaker System

5. LOP-CQ-03, Sound Powered Telephone System

6. LOP-CQ-04, Intra-Plant Radio System

7. OP-AA-104-101, Communications October 2017 LS 2-102 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

4. 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 > 200ºF for > Table C2 duration.

Table C2 RCS Heat-up Duration Thresholds RCS Status Containment Closure Heat-up Duration Status Intact Not Applicable 60 minutes*

Established 20 minutes*

Not Intact Not Established 0 minutes

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, then EAL #1 is not 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.

October 2017 LS 2-103 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 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. NEI 99-01 Rev 6, CA3

2. Technical Specifications 3.6.1.1

3. Technical Specifications 3.6.4.1

4. OU-AA-103, Shutdown Safety

5. OU-LA-104, Shutdown Safety Management Program

6. LGP-1-S1, Master Startup Checklist

7. LGP-1-1, Normal Unit Startup

8. LOR 1(2)H13-P601-C204, RHR Shutdown Cooling Line High Temperature October 2017 LS 2-104 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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 > 200ºF.

OR

2. Loss of the following for > 15 minutes.

ALL RCS temperature indications AND

ALL RPV water level indications Basis:

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

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

This IC addresses an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA5.

RCS is intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown mode of operation (e.g. no freeze seals, or steam line nozzle plugs, etc.).

October 2017 LS 2-105 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown.

EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA6 based on an inventory loss or IC CA5 based on exceeding plant configuration-specific time criteria.

Basis Reference(s):

1. NEI 99-01 Rev 6, CU3

2. Technical Specifications Table 1.1-1

3. LGP-1-S1, Master Startup Checklist

4. LGP-1-1, Normal Unit Startup

5. LGA-001, RPV Control

6. LPGP-PSTG-01S03 Plant Specific Technical Guidelines Section 3 - Cautions

7. LOP-SF-06, Filling the Reactor, Reactor Well and Dryer/Separator Pit Through Feedwater with Suppression Pool Cleanup

8. LOR 1(2)H13-P601-C204, RHR Shutdown Cooling Line High Temperature October 2017 LS 2-106 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. a. RPV Water Level < -161 inches (TAF) for > 30 minutes.

AND

b. Any Containment Challenge Indication (Table C4)

OR

2. a. RPV water level cannot be monitored for > 30 minutes.

AND

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

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

OR

Refuel floor Rad monitor 0D21-K604A >3000 mR/hr.

AND

c. ANY Containment Challenge Indication (Table C4)

Table C3 Indications of RCS Leakage

UNPLANNED floor or equipment sump level rise*

UNPLANNED Suppression Pool level rise*

UNPLANNED vessel make up rate rise*

Observation of leakage or inventory loss

Rise in level is attributed to a loss of RPV inventory.

October 2017 LS 2-107 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 Emergency Action Level (EAL) (cont):

Table C4 Containment Challenge Indications

Primary Containment Hydrogen Concentration > 6% and Oxygen > 5%

UNPLANNED rise in containment pressure

CONTAINMENT CLOSURE not established*

ANY Secondary Containment radiation monitor > LGA-002 Maximum Safe operating level.

if CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to a General Emergency is not required.

Basis:

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

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

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

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guidelines (PAG) exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

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.

October 2017 LS 2-108 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CG6 (cont)

Basis (cont):

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.

In EAL #2.a the 30-minute criterion is tied to a readily recognizable event start time (i.e.,

the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Basis Reference(s):

1. NEI 99-01 Rev 6, CG1

2. LGA-001, RPV Control

3. Technical Specifications 3.6.1.1

4. Technical Specifications 3.6.4.1

5. LGA-003, Primary Containment Control

6. LGA-011, Hydrogen Control

7. LaSalle PSTG Section 5B, Hydrogen Control

8. LGA-002, Secondary Containment Control

9. UFSAR 3.6.2

10. LIS-NR-301(401), Unit 1(2) Source Range Monitor Rod Block Functional Test October 2017 LS 2-109 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. With CONTAINMENT CLOSURE not established, RPV water level

< - 147 inches (Level 1)

OR

2. With CONTAINMENT CLOSURE established, RPV water level < - 161 inches (TAF)

OR

3. a. RPV water level cannot be monitored for > 30 minutes AND

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

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

OR

Refuel Floor Rad monitor 0D21-K604A >3000 mR/hr.

Table C3 Indications of RCS Leakage

UNPLANNED floor or equipment sump level rise*

UNPLANNED Suppression pool level rise*

UNPLANNED vessel make up rate rise*

Observation of leakage or inventory loss

Rise in level is attributed to a loss of RPV inventory.

October 2017 LS 2-110 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)

Basis:

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

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

The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs #1 and #2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

In EAL #3.a the 30-minute criterion is tied to a readily recognizable event start time (i.e.,

the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV water level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or 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.

October 2017 LS 2-111 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CS6 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, CS1

2. LGA-001, RPV Control

3. Technical Specifications Table 3.3.5.1-1

4. LOP-SF-06, Filling the Reactor, Reactor Well and Dryer/Separator Pit Through Feedwater with Suppression Pool Cleanup

5. LGA-001, RPV Control

6. UFSAR 5.2.5

7. LOP-NB-03, Troubleshooting Drywell Leakage

8. LIS-NR-301(401), Unit 1(2) Source Range Monitor Rod Block Functional Test

9. LGP-1-1, Normal Unit Startup

10. LGP-1-S1, Master Startup Checklist October 2017 LS 2-112 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. Loss of RPV inventory as indicated by level < - 83 inches (Level 2).

OR

2. a. RPV water level cannot be monitored for > 15 minutes.

AND

b. Loss of RPV inventory per Table C3 indications.

Table C3 Indications of RCS Leakage

UNPLANNED floor or equipment sump level rise*

UNPLANNED Suppression pool level rise*

UNPLANNED vessel make up rate rise*

Observation of leakage or inventory loss

Rise in level is attributed to a loss of RPV inventory.

Basis:

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

This IC addresses 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.

October 2017 LS 2-113 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CA6 (cont)

Basis (cont):

EAL #1 Basis A lowering of water level below -83 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. NEI 99-01 Rev 6, CA1

2. LOP-SF-06, Filling the Reactor, Reactor Well and Dryer/Separator Pit Through Feedwater with Suppression Pool Cleanup

3. LGA-001, RPV Control

4. Technical Specifications Table 3.3.5.1-1

5. LGA-001, RPV Control

6. UFSAR 5.2.5

7. LOP-NB-03, Troubleshooting Drywell Leakage October 2017 LS 2-114 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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 floor or equipment sump level rise*

UNPLANNED Suppression pool level rise*

UNPLANNED vessel make up rate rise*

Observation of leakage or inventory loss

Rise in level is attributed to a loss of RPV inventory.

October 2017 LS 2-115 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

EAL #1 recognizes that the minimum required RPV water level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL #2 addresses a condition where all means to determine RPV water level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA6 or CA5.

October 2017 LS 2-116 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTIONS CU6 (cont)

Basis Reference(s):

1. NEI 99-01, Rev. 6 CU1

2. LOP-SF-06, Filling the Reactor, Reactor Well and Dryer/Separator Pit Through Feedwater with Suppression Pool Cleanup

3. LGA-001, RPV Control

4. Technical Specifications Table 3.3.1.1-1.

5. LPGP-CALC-2

6. UFSAR 5.2.5 October 2017 LS 2-117 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 Initiating Condition:

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

Operating Mode Applicability:

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

1. A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

AND

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

OR

b. Damage to spent fuel has occurred or is IMMINENT Table H1 Safety Functions

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

RPV Water Level (ability to cool the core)

RCS Heat Removal (ability to maintain heat sink)

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

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

October 2017 LS 2-118 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 (cont)

Basis (cont):

HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to

1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

Basis Reference(s):

1. NEI 99-01, Rev. 6 HG1

2. Station Security Plan - Appendix C October 2017 LS 2-119 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition:

HOSTILE ACTION within the PROTECTED AREA.

Operating Mode Applicability:

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

A notification from the Security Force that a HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

October 2017 LS 2-120 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 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 HG1.

Basis Reference(s):

1. NEI 99-01 Rev 6, HS1

2. Station Security Plan - Appendix C October 2017 LS 2-121 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 Initiating Condition:

HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

Operating Mode Applicability:

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

1. A validated notification from NRC of an aircraft attack threat < 30 minutes from the site.

OR

2. Notification by the Security Force that a HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA.

Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

OWNER CONTROLLED AREA (OCA): The property associated with the station and owned by the company. Access is normally limited to persons entering for official business.

PROTECTED AREA: An area that normally encompasses all controlled areas within the security protected area fence.

HOSTILE FORCE: Any individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

October 2017 LS 2-122 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)

Basis (cont):

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE.

Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

EAL #1 Basis Addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with LOA-SY-001, Security Abnormal Procedure.

EAL #2 Basis Is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Escalation of the emergency classification level would be via IC HS1.

October 2017 LS 2-123 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 (cont)

Basis Reference(s):

1. NEI 99-01 Rev 6, HA1

2. Station Security Plan - Appendix C

3. LOA-SY-001, Security Abnormal Procedure October 2017 LS 2-124 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition:

Confirmed SECURITY CONDITION or threat.

Operating Mode Applicability:

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

1. Notification of a credible security threat directed at the site as determined per SY-AA-101-132, Security Assessment and Response to Unusual Activities.

OR

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

OR

3. Notification by the Security Force of a SECURITY CONDITION that does not involve a HOSTILE ACTION.

Basis:

SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety.

Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1.

October 2017 LS 2-125 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 (cont)

Basis (cont):

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

EAL #1 Basis Addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with SY-AA-101-132.

EAL #2 Basis Addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with LOA-SY-001, Security Abnormal Procedure.

EAL #3 Basis References Security Force because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

Escalation of the emergency classification level would be via IC HA1.

Basis Reference(s):

1. NEI 99-01 Rev 6, HU1

2. SY-AA-101-132, Security Assessment and Response to Unusual Activities

3. Station Security Plan - Appendix C

4. NRC Safeguards Advisory 10/6/01

5. Letter from Mr. B. A. Boger (NRC) to Ms. Lynette Hendricks (NEI) dated 2/4/02

6. LOA-SY-001, Security Abnormal Procedure October 2017 LS 2-126 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

1. A Control Room evacuation has resulted in plant control being transferred from the Control Room to alternate locations per:

LOA-RX-101(201), Unit 1(2) Control Room Evacuation Abnormal OR

LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Fire in the Control Room OR AEER.

AND

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

Table H1 Safety Functions

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

RPV Water Level (ability to cool the core)

RCS Heat Removal (ability to maintain heat sink)

Basis:

The time period to establish control of the plant starts when either:

a. Control of the plant is no longer maintained in the Main Control Room OR

b. The last Operator has left the Main Control Room.

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plan control to alternate locations is a precursor to a challenge to any fission product barriers within a relatively short period of time.

October 2017 LS 2-127 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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. NEI 99-01, Rev 6 HS6

2. LOA-RX-101(201), Unit 1(2) Control room Evacuation Abnormal

3. LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Fire in the Control Room OR AEER October 2017 LS 2-128 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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:

LOA-RX-101(201), Unit 1(2) Control Room Evacuation Abnormal OR

LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Fire in the Control Room OR AEER.

Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel.

Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Escalation of the emergency classification level would be via IC HS2.

Basis Reference(s):

1. NEI 99-01, Rev 6 HA6

2. LOA-RX-101(201), Unit 1(2) Control Room Evacuation Abnormal

3. LOA-FX-101 (201), Unit 1(2) Safe Shutdown with a Fire in the Control Room OR AEER October 2017 LS 2-129 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time.

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

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

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

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

Field verification of a single fire alarm Table H2 Vital Areas

Reactor Building (when inerted the Drywell is exempt)

Control Room

Auxiliary Building

Unit and Shared Emergency Diesel Generator Rooms

Switchgear and Battery Rooms

Remote Shutdown Rooms

CSCS Pump Rooms

LSH (for 0E12-F300 access only)

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 October 2017 LS 2-130 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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.

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.

October 2017 LS 2-131 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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.

October 2017 LS 2-132 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 (cont)

Basis (cont):

Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA2 or MA5.

Basis Reference(s):

1. NEI 99-01, Rev 6 HU4

2. UFSAR 3.8

3. LOA-FP-101(201), Fire Protection System Abnormal October 2017 LS 2-133 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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:

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

For emergency classification if EAL # 2.b is not able to be confirmed, then the occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director in < 15 mins of the event.

1. Seismic event > Operating Basis Earthquake (OBE) as indicated by:

Alarm on Panel 1PM10J B503 SEISMIC OBE/SSE LEVEL EXCEEDED OR

OBE LED lit on 0PA11J 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 mins of the event:

The earthquake resulted in Modified Mercalli Intensity (MMI) > VI and occurred

< 3.5 miles of the plant.

The earthquake was magnitude > 6.0

The earthquake was magnitude > 5.0 and occurred < 125 miles of the plant.

Basis:

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)1. An earthquake greater 1

An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional.

October 2017 LS 2-134 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 (cont)

Basis (cont):

than an OBE but less than a Safe Shutdown Earthquake (SSE)2 should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections).

Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are in excess of 0.08g).

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.b and the accompanying note is included to ensure that a declaration does not result from felt vibrations caused by a non-seismic source (e.g., a dropped load). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., call to USGS, check internet source, etc.) however, the verification action must not preclude a timely emergency declaration. This guidance recognizes that it may cause the site to declare an Unusual Event while another site, similarly affected but with readily available OBE indications in the Control Room, may not.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA2 or MA5.

Basis Reference(s):

1. NEI 99-01, Rev 6 HU2

2. UFSAR 3.3

3. UFSAR 3.4

4. UFSAR 3.7

5. UFSAR 3.8

6. UFSAR 3.11

7. LOR-1PM10J-B503 Seismic Operating Basis Earthquake (OBE)/Safe Shutdown Earthquake (SSE) Level Exceeded

8. US NRC Reg. Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Earthquake Actions.

2 An SSE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.

October 2017 LS 2-135 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 Initiating Condition:

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

Operating Mode Applicability:

3, 4 Emergency Action Level (EAL):

Note:

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

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

Table H3 Areas with Entry Related Mode Applicability Entry Related Mode Area Applicability Reactor Building 673 SE Corner Room 694 NW Corner Room SE Corner Room Raceway Mode 3 and 4 740 NE Valve Room NE SW 761 NE SW Auxiliary Building 674 CSCS Rooms Mode 3 and 4 731 AEER 736 DG Penthouse 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.

October 2017 LS 2-136 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)

Basis (cont):

Assuming all plant equipment is operating as designed, normal operation is capable from the Main Control Room (MCR). The plant is also able to transition into a hot shutdown condition from the MCR, therefore Table H3 is a list of plant rooms or areas with entry-related mode applicability that contain equipment which require a manual/local action necessary to transition the plant from normal plant operation to cooldown and shutdown as specified in normal operating procedures (establish shutdown cooling), where 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 Directors judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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

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

The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

October 2017 LS 2-137 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 (cont)

Basis (cont):

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that generate smoke, that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

The Operating Mode Applicability of this EAL has been revised from All Modes to modes 3 and 4 due to the mode applicability of the areas of concern in Table H-3. In the future should the areas of concern in Table H-3 be revised then the Operating Mode Applicability of this EAL should be reevaluated.

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

Basis Reference(s):

1. NEI 99-01, Rev 6 HA5

2. UFSAR 9.5

3. OP-AA-106-103 Chemical Release Or Spill Assessment And Response Determination

4. OP-LA-106-103 Hazardous Materials (Hazmat) Release Supplemental Information

5. EN-AA-702 Chemical Non-Emergency Response

6. ACIT 660892-16, Station Halon Discharge IDLH Evaluation October 2017 LS 2-138 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 Initiating Condition:

Hazardous Event Operating Mode Applicability:

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

Note:

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

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

1. Tornado strike within the PROTECTED AREA.

OR

2. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical specifications for the current operating mode.

OR

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

OR

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

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.

EAL #1 Basis Addresses a tornado striking (touching down) within the Protected Area.

October 2017 LS 2-139 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)

Basis (cont):

EAL #2 Basis Addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

EAL #3 Basis Addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.

EAL #4 Basis Addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended 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.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M, H or C.

October 2017 LS 2-140 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU6 (cont)

Basis Reference(s):

1. NEI 99-01, Rev 6 HU3

2. UFSAR 3.3

3. UFSAR 3.4

4. UFSAR 3.7

5. UFSAR 3.8

6. UFSAR 3.11

7. LOA-TORN-001, High Winds/Tornado

8. Drawing S-01A, Composite Site Plan

9. LOA-FLD-001, Flooding

10. Drawing M-24, Flood Plan October 2017 LS 2-141 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-142 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 HS7 October 2017 LS 2-143 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 HA7 October 2017 LS 2-144 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 October 2017 LS 2-145 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle 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 exceeding ANY Table E1 on-contact radiation reading.

Table E1 On-contact radiation readings Cask Type Location of on-contact radiation reading Top of the Side of the spent fuel spent fuel cask, excluding inlet and HI-STORM cask outlet ducts (mrem/hr) (mrem/hr)

Labeled as

> 40 > 220 XXX-A3 Labeled as

> 60 > 600 XXX-A8 Basis:

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

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) : A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The word cask, as used in this EAL, refers to the storage container in use at the site for dry storage of irradiated fuel. The issues of concern are the creation of a potential or actual release path to the environment, degradation of any fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of damage is determined by radiological survey. The technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The October 2017 LS 2-146 EP-AA-1005 Addendum 3 (Revision 3)

LaSalle Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTIONS E-HU1 (cont)

Basis (cont):

emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Basis Reference(s):

1. NEI 99-01, Rev 6 E-HU1

2. Certificate of Compliance No.1014, Amendment No. 3 and 8 Revision 1 Appendix A October 2017 LS 2-147 EP-AA-1005 Addendum 3 (Revision 3)

RS-17-162, Radiological Emergency Plan Addendum Revision

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