View source for L-2019-111, Attachments 1 Through 5: Evaluation of Proposed Changes, EAL Technical Bases Document, Mark-Up of EAL, EAL Comparison Matrix and Emergency Action Level Matrix Wall Charts

{{Adams
| number = ML19275G849
| issue date = 09/30/2019
| title = Attachments 1 Through 5: Evaluation of Proposed Changes, EAL Technical Bases Document, Mark-Up of EAL, EAL Comparison Matrix and Emergency Action Level Matrix Wall Charts
| author name = 
| author affiliation = Florida Power &amp; Light Co
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000335, 05000389
| license number = 
| contact person = 
| case reference number = L-2019-111
| package number = ML19275G789
| document type = Emergency Preparedness-Emergency Plan
| page count = 615
}}

=Text=
{{#Wiki_filter:L-2019-111 Attachment 1 Page 1 of 8 ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES

==Subject:==
License Amendment Request Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" Table of Contents 1.0    

==SUMMARY==
DESCRIPTION ................................................................................................................... 2 2.0    DETAILED ESCRIPTION........................................................................................................................ 2 2.1    Proposed Initiating Conditions (ICs) and EALs ............................................................................... 2 2.2    Deviations and Differences ............................................................................................................ 2 2.3    Generic Differences........................................................................................................................ 3 2.4    Operational Modes ........................................................................................................................ 3 2.5    Instrumentation Used for EALs ...................................................................................................... 4 2.6    Background Technical Information ................................................................................................ 4

==3.0    TECHNICAL EVALUATION==
 ................................................................................................................... 4

==4.0    REGULATORY EVALUATION==
................................................................................................................ 5 4.1    Applicable Regulatory Requirements / Criteria ............................................................................. 5 4.2    No Significant Hazards Consideration ............................................................................................ 6 4.3    Conclusion ...................................................................................................................................... 7

==5.0    ENVIRONMENTAL CONSIDERATION==
 ................................................................................................. 7

==6.0    REFERENCES==
 ....................................................................................................................................... 8

L-2019-111 Attachment 1 Page 2 of 8 1.0     

==SUMMARY==
DESCRIPTION Florida Power & Light Company (FPL) proposes to revise the St Lucie currently approved Emergency Plan (EP) Emergency Action Level (EAL) scheme, which is based on the Nuclear Energy lnstitute's (NEl's) guidance established in NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels" [Reference 1]. NextEra is proposing to adopt the EAL schemes based on the guidance provided in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC [Reference 2].
2.0     DETAILED ESCRIPTION The proposed changes involve revising St Lucies EAL scheme, which is currently based on NEI 99-01, Revision 5, to a scheme based on NEI 99-01, Revision 6. Enhancements provided by revision 6 of the guidance include:
: 1. Clarification of numerous EALs that have been typically misinterpreted by the industry in the development of their site-specific EAL scheme,
: 2. Clarification of the intent of EALs that have been historically misclassified,
: 3. Providing additional guidance for the development of EALs for current non-passive reactor designs as well as possible future reactor designs that are non-passive,
: 4. Incorporating lessons learned from industry events (i.e., Fukushima and others) and NUREG/CR-7154, Risk Informing Emergency Preparedness Oversight: Evaluation of Emergency Action Levels - A Pilot Study of Peach Bottom, Surry and Sequoyah, and
: 5. A detailed review of the guidance to re-validate that the EALs are appropriate and are at the necessary emergency classification level based upon 32 years of industry and NRC experience with EAL scheme development and implementation.
2.1     Proposed Initiating Conditions (ICs) and EALs , St Lucie Emergency Action Levels Bases Document, provides a markup of NEI 99-01, Revision 6. The markup shows the St Lucie-specific proposed changes to NEI 99-01. Attachment 2 provides a clean copy of St Lucie Emergency Action Levels Bases Document.
2.2     Deviations and Differences  contains a matrix that provides in tabular format a comparison of the initiating conditions (ICs) and EALs in NEI 99-01 to the ICs and EALs proposed for St Lucie. The comparison evaluates differences and deviations in accordance with the guidance in Regulatory Issue Summary (RIS) RIS 2003-18, Use of NEI 99-01, "Methodology for Development of Emergency Action Levels,"
Revision 4, Dated January 2003, Supplements 1 and 2 [References 3, 4] and provides the basis for each difference and deviation.
As discussed in RIS 2003-18, Supplement 1, differences and deviations are defined as:
* A difference is an EAL change where the basis scheme guidance differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the site-specific proposed EAL. Examples of

L-2019-111 Attachment 1 Page 3 of 8 differences include the use of site-specific terminology or administrative re-formatting of site-specific EALs.
* A deviation is an EAL change where the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the site-specific proposed EAL. Examples of deviations include the use of altered mode applicability, altering key words or time limits, or changing words of physical reference (protected area, safety-related equipment, etc.).
A number of differences are the result of adding plant-specific information to the EALs.
2.3     Generic Differences The differences below apply throughout the set of EALs and are not specifically identified in each instance in the comparison matrix as a difference.
NEI 99-01, Rev 6 EALs                                 St Lucie EALs References BWRs                                       Deleted BWR references as appropriate Uses A for radiological effluent/radiation level      Uses R for radiological effluent/radiation level ICs                                                   ICs Uses E-HU for ISFSI ICs                               Uses EU for ISFSI ICs Emergency Classification ICs are presented in         Emergency Classification ICs are presented in ascending order (NOUE - GE)                          descending order (GE - NOUE) 2.4     Operational Modes Mode applicability of the proposed ICs and EALs is consistent with the NEI 99-01, Rev.
: 6. The following table provides the operating modes for St. Lucie defined by the Technical Specifications (TS).

L-2019-111 Attachment 1 Page 4 of 8 THERMAL MODE                   K eff     POWER
* TAVG
: 1. Power Operation                0.99       > 5%                   325&deg;F
: 2. Startup                        0.99        5%                   325&deg;F
: 3. Hot Standby                   < 0.99         0                    325&deg;F
: 4. Hot Shutdown                  < 0.99         0             > 200&deg;F & < 325&deg;F
: 5. Cold Shutdown                 < 0.99         0                    200&deg;F
: 6. Refueling **                   0.95         0                    140&deg;F Defueled ***                    NA          NA           NA - no fuel in reactor vessel
         *Excluding decay heat
         **Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed
         ***Not defined in the TS In addition to the TS defined operational modes, NEI 99-01, Revision 6, defines the following additional mode:
Defueled: All reactor fuel removed from the reactor vessel (i.e., full core off load during refueling or extended outage).
St Lucie procedures recognize and are consistent with the definition of a defueled condition.
2.5      Instrumentation Used for EALs FPL reviewed the specified values used as EAL setpoints and verified they are within the calibrated range of the referenced instrumentation.
2.6      Background Technical Information  contains technical information, such as plant drawings, calculations etc., that support the proposed EALs. Attachment 2, Markup of St Lucie Emergency Action Levels - Initiating Conditions, Threshold Values and Basis; and Attachment 4, Deviations and Differences, provide a reference to the specific document in Attachment 5 containing technical information for each EAL as applicable.

==3.0      TECHNICAL EVALUATION==

FPL has evaluated the proposed changes to determine whether applicable regulations and requirements have been met. NEI 99-01 guidance methodology includes many years of development along with use and implementation. The guidance has been subject to NRC reviews and approval.
The St Lucie EAL scheme currently in place is based on the methodology in NEI 99-01, Revision 5.
NEI 99-01, Revision 6, is the latest guidance endorsed by the NRC and provides guidance to nuclear power plant operators for the development of a site-specific emergency action level scheme.

L-2019-111 Attachment 1 Page 5 of 8 10 CFR 50.47(b)(4) requires that emergency plans include a standard emergency classification and action level scheme. This scheme is a fundamental component of an emergency plan in that it provides the defined thresholds that will allow site personnel to rapidly implement a range of pre-planned emergency response measures. An emergency classification scheme also facilitates timely decision-making by an offsite response organization concerning the implementation of precautionary or protective actions for the public.
NEI 99-01, Revision 6, contains a generic set of ICs, EALs, and fission product barrier status thresholds. It also includes supporting technical basis information, developer notes, and recommended classification instructions for users. The methodology described in this document is consistent with NRC requirements and guidance. In particular, this methodology was specifically endorsed by the NRC in a March 28, 2013, letter from NRC to NEI [Reference 5] and determined to provide an acceptable approach in meeting requirements of 10 CFR 50.47(b)(4), applicable requirements of 10 CFR 50, Appendix E, and the associated planning standard evaluation elements contained in NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,"
November 1980.
10 CFR 50, Appendix E, Section IV.8.2 requires that a licensee desiring to change its entire EAL scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. The proposed change to the St Lucie EAL scheme from NEI 99-01 Revision 5 to NEI 99-01 Revision 6 guidance does not reduce the capability to meet the applicable emergency planning standards and requirements in 10 CFR 50.47(b) and 10 CFR 50, Appendix E.
Accordingly, pursuant to the requirements of 10 CFR 50, Appendix E, Section IV.8.2, NextEra requests NRC review and approval of the proposed changes to the EAL scheme as a license amendment request in accordance with 10 CFR 50.90.

==4.0      REGULATORY EVALUATION==

4.1      Applicable Regulatory Requirements / Criteria The regulations in 10 CFR 50.54(q) provide direction to licensees seeking to revise their emergency plan. The requirements related to nuclear power plant emergency plans are contained in the standards in 10 CFR 50.47, "Emergency Plans," and the requirements of Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities."
Paragraph 10 CFR 50.47(a)(1) says that no operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Section 50.47(b) contains standards that onsite and offsite emergency response plans must meet for the NRC staff to make a positive finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. One of these standards, 10 CFR 50.47(b)(4), requires that emergency plans include a standard emergency classification and action level scheme.
10 CFR 50, Appendix E, Section IV.B, "Assessment Actions," requires that emergency plans include emergency action levels (EALs) that are to be used as criteria for determining the need for notification and participation of local and state agencies, and for determining when and what type of protective measures should be considered to protect the health and safety of individuals both onsite and offsite. EALs are to be based on plant conditions and instrumentation, as well as onsite and offsite radiological monitoring. Section IV.B provides that initial EALs shall be discussed and agreed on by the applicant and state and local authorities, be approved by the NRC, and reviewed annually

L-2019-111 Attachment 1 Page 6 of 8 thereafter with state and local authorities. Therefore, a revision to EALs will require NRC approval prior to implementation if it involves (1) changing from one EAL scheme to another (e.g., NEI 99-01 Rev. 5 to NEI 99-01 Rev. 6), (2) proposing an alternate method to comply with the regulations, or (3) the EAL revision proposed by the licensee decreases the effectiveness of the emergency plan.
NRC Regulatory Issue Summary (RIS) 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes", issued April 19, 2011, says that a change in an EAL scheme to incorporate the improvements provided in NUMARC/NESP-007 or NEI 99-01 would not decrease the overall effectiveness of the emergency plan, but due to the potential safety significance of the change, the change needs prior NRC review and approval.
The proposed changes meet the above regulatory requirements.
4.2      No Significant Hazards Consideration FPL proposes to revise St Lucies currently approved Emergency Plan (EP) Emergency Action Level (EAL) scheme, which is based on the Nuclear Energy lnstitute's (NEl's) guidance established in NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels." FPL is proposing to adopt the EAL schemes based on the guidance provided in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC.
FPL has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10CFR50.92, "Issuance of Amendment,"
as discussed below.
: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes to the St Lucie emergency action levels neither involve any physical changes to plant equipment or systems nor do they alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors, and they do not alter design assumptions, plant configuration, or the manner in which the plant is operated and maintained. The proposed change does not adversely affect the ability of structures, systems or components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant SSC. No physical changes are made to the plant, and emergency action levels are not accident initiators; so no new causal mechanisms are introduced.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

L-2019-111 Attachment 1 Page 7 of 8
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not impact operation of the plant and no accident analyses are affected by the proposed changes. The changes do not affect the Technical Specifications or the method of operating the plant. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes.
The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.
Therefore, the proposed changes do not involve a significant reduction in the margin of safety.
4.3      Conclusion In conclusion, based on the considerations above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the general public.

==5.0      ENVIRONMENTAL CONSIDERATION==

The proposed change is applicable to emergency planning requirements involving the proposed adoption of the NRC-endorsed EAL guidance as described in NEI 99-01, Revision 6, and does not reduce the capability to meet the emergency planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR 50, Appendix E. The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 because the amendment approves an acceptable EAL scheme that is required for operation of the facility. The proposed change does not involve (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite; or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is required in connection with the proposed change.

L-2019-111 Attachment 1 Page 8 of 8

==6.0    REFERENCES==
: 1. Nuclear Energy Institute, NEI 99-01 Revision 5, "Methodology for Development of Emergency Action Levels," February 2008 (ADAMS Accession No. ML080450149)
: 2. Nuclear Energy Institute (NEI) 99-01, Revision 6, "Methodology for Development of Emergency Action Levels for Non Passive Reactors," November 2012 (ADAMS Accession No. ML12326A805)
: 3. NRC Regulatory Issue Summary 2003-18, Supplement 1: Use of Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, Dated January 2003, July 13, 2004
: 4. NRC Regulatory Issue Summary 2003-18, Supplement 2: Use of Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, Dated January 2003, December 12, 2005
: 5. NRC letter "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99- 01, Revision 6, Dated November, 2012 (TAC No. D92368)," March 28, 2013 (ADAMS Accession No. ML12346A463)

St. Lucie Plant                                         L-2019-111 Emergency Action Level Technical Bases Document       Attachment 2 Emergency Action Level Technical Bases Document

St. Lucie Plant Emergency Action Level Technical Bases Document Table of Contents

==1.0   INTRODUCTION==
 ....................................................................................................... 3 2.0   DISCUSSION............................................................................................................. 3 2.1   Background ...................................................................................................... 3 2.2   Fission Product Barriers ................................................................................... 4 2.3   Fission Product Barrier Classification Criteria .................................................. 4 2.4   EAL Organization ............................................................................................. 4 2.5   Technical Bases Information ............................................................................ 7 2.6   Operational Mode Applicability ......................................................................... 8 3.0   GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................. 9 3.1   General Considerations .................................................................................... 9 3.2   Classification Methodology ............................................................................. 10

==4.0   REFERENCES==
 ........................................................................................................ 14 4.1   Developmental................................................................................................ 14 4.2   Implementing .................................................................................................. 14 5.0   DEFINITIONS, ACRONYMS & ABBREVIATIONS .................................................. 15 5.1   Definitions ....................................................................................................... 15 5.2   Abbreviations/Acronyms ................................................................................. 20 6.0   PSL-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE........................................... 23 7.0   ATTACHMENTS ...................................................................................................... 27 7.1   Attachment 1, Emergency Action Level Technical Bases............................... 27 7.2   Attachment 2, Safe Operation & Shutdown Areas Tables R-5 & H-2 Bases .. 27 Category R - Abnormal Rad Levels / Rad Effluent............................................................. 28 Category C - Cold Shutdown / Refueling System Malfunction ........................................... 71 Category E - Independent Spent Fuel Storage Installation (ISFSI).................................. 105 Category F - Fission Product Barrier Degradation ........................................................... 108 Category H - Hazards and Other Conditions Affecting Plant Safety ................................ 155 Category S - System Malfunction .................................................................................... 191 Page 2 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document

==1.0    INTRODUCTION==

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the NEI 99-1 Revision 6 EAL Upgrade Project for St. Lucie Plant (PSL). It should be used to facilitate review of the PSL EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EPIP-01, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation.
This information may assist the Emergency Coordinator (EC) in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.
The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.
Because the information in a basis document can affect emergency classification decision-making (e.g., the EC refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).
2.0    DISCUSSION 2.1    Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the Saint Lucie Plant Emergency Plan.
In 1992, the NRC endorsed NUMARC/NESP-007 Methodology for Development of Emergency Action Levels as an alternative guidance to the original Standard Review Plan and NUREG-0654 EAL schemes.
NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:
* Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
* Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).
* Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).
Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ref. 4.1.1), PSL conducted an EAL implementation upgrade project that produced the EALs discussed herein.
Page 3 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 2.2     Fission Product Barriers Fission product barrier 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.
Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies a greater probability of barrier loss and reduced certainty of maintaining the barrier.
The primary fission product barriers are:
A. Fuel Clad Barrier (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System Barrier (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment Barrier (CTMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.
2.3     Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriers and loss or potential loss of the third barrier 2.4     EAL Organization The PSL EAL scheme includes the following features:
* Division of the EAL set into three broad groups:
Page 4 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document o EALs applicable under any plant operational modes - This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operational modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.
o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.
The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
* Within each group, assignment of EALs to categories and subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The PSL EAL categories are aligned to and represent the NEI 99-01 Recognition Categories. Subcategories are used in the PSL scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The PSL EAL categories and subcategories are listed below.
The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.
Page 5 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document EAL Groups, Categories and Subcategories EAL Group/Category                                EAL Subcategory Any Operating Mode:
R - Abnormal Rad Levels / Rad Effluent    1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions          1 - Security Affecting Plant Safety                2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - EC Judgment E - Independent Spent Fuel Storage        1 - Confinement Boundary Installation (ISFSI)
Hot Conditions:
S - System Malfunction                    1 - Loss of Emergency AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation   None Cold Conditions:
C - Cold Shutdown / Refueling System      1 - RCS Level Malfunction                           2 - Loss of Emergency AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems Page 6 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 2.5      Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:
Category Letter & Title Subcategory Number & Title Initiating Condition (IC)
Site-specific description of the generic IC given in NEI 99-01 Rev. 6.
EAL Identifier (enclosed in rectangle)
Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:
: 1. First character (letter): Corresponds to the EAL category as described above (R, C, E, F, H or S)
: 2. Second character (letter): The emergency classification (G, S, A or U)
G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event
: 3. Third character (number): Subcategory number within the given category.
Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).
: 4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).
Classification (enclosed in rectangle):
General Emergency (G), Site Area Emergency (S), Alert (A) or Unusual Event (U).
EAL (enclosed in rectangle)
Exact wording of the EAL as it appears in the EAL Classification Matrix.
Page 7 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 -
Cold Shutdown, 6 - Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitions)
Definitions:
If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.
Basis:
An EAL basis section that provides PSL-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.
Reference(s):
Source documentation from which the EAL is derived 2.6      Operational Mode Applicability THERMAL MODE                   K eff       POWER
* TAVG
: 1. Power Operation               0.99         > 5%                  325&deg;F
: 2. Startup                       0.99          5%                  325&deg;F
: 3. Hot Standby                  < 0.99           0                   325&deg;F
: 4. Hot Shutdown                 < 0.99           0            > 200&deg;F & < 325&deg;F
: 5. Cold Shutdown                < 0.99           0                   200&deg;F
: 6. Refueling **                  0.95           0                   140&deg;F Defueled                       NA            NA          NA - no fuel in reactor vessel
* Excluding Decay Heat
               **  Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
Page 8 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 3.0      GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1      General Considerations When making an emergency classification, the EC must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operational Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier thresholds.
3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8).
3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the EC should not wait until the applicable time has elapsed, but 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 radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
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St. Lucie Plant Emergency Action Level Technical Bases Document 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10CFR 50.72 (ref.
4.1.4).
3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording 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 threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
3.1.6 Emergency Coordinator (EC) Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the EC with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The EC will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.
3.2     Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process clock started.
When assessing an EAL that specifies a time duration for the potentially classifiable condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref.
4.1.8).
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St. Lucie Plant Emergency Action Level Technical Bases Document 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:
* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency should be declared.
There is no additive effect from multiple EALs meeting the same ECL. For example:
* If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).
3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.
3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the EC must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the EC, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
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St. Lucie Plant Emergency Action Level Technical Bases Document 3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).
3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram.
3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response - In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:
An ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.
It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event.
Emergency classification assessments must be deliberate and timely, with no undue delays.
The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the EC completing the review Page 12 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.
In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10CFR 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).
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St. Lucie Plant Emergency Action Level Technical Bases Document

==4.0   REFERENCES==

4.1   Developmental 4.1.1    NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2    RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.
4.1.3    NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4    10 CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5    10 CFR 50.73 License Event Report System 4.1.6    Technical Specifications Table 1.2, Operational Modes 4.1.7    C-200 Offsite Dose Calculation Manual (ODCM) 4.1.8    NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9    St. Lucie Plant Radiological Emergency Plan 4.1.10 Certificate of Compliance Appendix A NUHOMS HD System Generic Technical Specifications 4.1.11 UFSAR Chapter 9.1 Fuel Storage and Handling 4.1.12 1(2)-GMM-68.02 Emergency Closure of Containment Penetrations, Personnel Hatch and Equipment Hatch 4.2   Implementing 4.2.1 EPIP-01, Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to PSL EAL Comparison Matrix 4.2.3 PSL EAL Matrix Page 14 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 5.0     DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1     Definitions (ref. 4.1.1 except as noted)
Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.
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.
CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC) (ref. 4.1.10).
CONTAINMENT CLOSURE The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist (ref.
4.1.12):
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
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.
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)
EXPLOSION Page 15 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 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 require a post-event inspection to determine if the attributes of an explosion are present.
FAULTED The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
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.
FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
FLOODING A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
GENERAL EMERGENCY 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 PAG exposure levels offsite for more than the immediate site area.
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 PSL 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 PSL. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).
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St. Lucie Plant Emergency Action Level Technical Bases Document HOSTILE FORCE One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
IMMINENT The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
IMPEDE(D)
Personnel access to a room or area is hindered to an extent that 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).
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.
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.
OWNER CONTROLLED AREA (OCA)
That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL (ref. 4.1.9).
PROJECTILE An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency (ref. 4.1.9).
REFUELING PATHWAY Refueling cavity, fuel transfer canal, and spent fuel pool (SFP), but not including the reactor vessel, comprise the refueling pathway (ref. 4.1.12).
RUPTURED The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
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St. Lucie Plant Emergency Action Level Technical Bases Document 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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
SITE AREA EMERGENCY 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 PAG exposure levels beyond the SITE BOUNDARY.
SITE BOUNDARY That line beyond which the land or property is not owned, leased or otherwise controlled by PSL (ref. 4.1.7).
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.
UNUSUAL EVENT Events are in progress or have occurred which indicate a potential degradation in 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.
VALID An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct Page 18 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
VISIBLE DAMAGE Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
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St. Lucie Plant Emergency Action Level Technical Bases Document 5.2        Abbreviations/Acronyms
&deg;F ....................................................................................................... Degrees Fahrenheit
&deg; ........................................................................................................................... Degrees AC ....................................................................................................... Alternating Current AFW....................................................................................................Auxiliary Feedwater AOP ................................................................................. Abnormal Operating Procedure ARM.............................................................................................. Area Radiation Monitor ATWS ...................................................................... Anticipated Transient Without Scram CDE .......................................................................................Committed Dose Equivalent CET ............................................................................................ Core Exit Thermocouple CFR ..................................................................................... Code of Federal Regulations CIAS .....................................................................Containment Isolation Actuation Signal CPM..................................................................................................... Counts Per Minute CTMT............................................................................................................. Containment DEF .....................................................................................................................Defueled DBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current D/G ......................................................................................................... Diesel Generator EAL ............................................................................................. Emergency Action Level EC ............................................................................................... Emergency Coordinator ECCS............................................................................ Emergency Core Cooling System ECL ................................................................................. Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency FAA.................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FC ........................................................................................................... Fuel Clad Barrier FEMA............................................................... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency GPM ....................................................................................................Gallons Per Minute HR .............................................................................................................. Heat Removal Hr. .............................................................................................................................. Hour HSM.........................................................................................Horizontal Storage Module IC ......................................................................................................... Initiating Condition ICC ............................................................................................ Inadequate Core Cooling ISFSI............................................................ Independent Spent Fuel Storage Installation Page 20 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Keff ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPSI .................................................................................... Low Pressure Safety Injection LRW........................................................................................................ Liquid Radwaste LWR................................................................................................... Light Water Reactor MCB.................................................................................................... Main Control Board MG ........................................................................................................... Motor Generator Min. ......................................................................................................................... Minute MPH........................................................................................................... Miles Per Hour mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MW .................................................................................................................... Megawatt NEI .............................................................................................. Nuclear Energy Institute NEIC ................................................................... National Earthquake Information Center NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................ Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD................................................... North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event OBE ...................................................................................... Operating Basis Earthquake OCA ...............................................................................................Owner Controlled Area ODCM............................................................................ Off-site Dose Calculation Manual ORO ................................................................................. Offsite Response Organization OTC ...................................................................................... Once-Through Core Cooling PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PORV .................................................................................. Power Operated Relief Valve PSIG ................................................................................ Pounds per Square Inch Gauge PSL ............................................................................................................. St. Lucie Plant PTS........................................................................................ Pressurized Thermal Shock R ........................................................................................................................ Roentgen RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RPS ........................................................................................ Reactor Protection System RVLMS ..............................................................Reactor Vessel Level Monitoring System SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus Page 21 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document SFP.......................................................................................................... Spent Fuel Pool SFSC .................................................................................. Safety Function Status Check SG ......................................................................................................... Steam Generator SI .............................................................................................................. Safety Injection SIAS ............................................................................. Safety Injection Actuation System SM ............................................................................................................... Shift Manager SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SRV ..................................................................................................... Safety Relief Valve TC (T/C)...................................................................................................... Thermocouple TEDE ............................................................................... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TS ................................................................................................Technical Specifications TSC .......................................................................................... Technical Support Center UFSAR ...................................................................Updated Final Safety Analysis Report USGS ............................................................................ United States Geological Survey Page 22 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 6.0    PSL-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a PSL EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the PSL EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
PSL             NEI 99-01 Rev. 6 Example EAL              IC EAL RU1.1            AU1           1 RU1.2            AU1           2 RU1.3            AU1           3 RU2.1            AU2           1 RA1.1            AA1           1 RA1.2            AA1           2 RA1.3            AA1           3 RA1.4            AA1           4 RA2.1            AA2           1 RA2.2            AA2           2 RA2.3            AA2           3 RA3.1            AA3           1 RA3.2            AA3           2 RS1.1            AS1           1 RS1.2            AS1           2 RS1.3            AS1           3 RS2.1            AS2           1 RG1.1            AG1           1 RG1.2            AG1           2 RG1.3            AG1           3 RG2.1            AG2           1 Page 23 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document PSL         NEI 99-01 Rev. 6 Example EAL          IC EAL CU1.1        CU1           1 CU1.2        CU1           2 CU2.1        CU2           1 CU3.1        CU3           1 CU3.2        CU3           2 CU4.1        CU4           1 CU5.1        CU5        1, 2, 3 CA1.1        CA1           1 CA1.2        CA1           2 CA2.1        CA2           1 CA3.1        CA3         1, 2 CA6.1        CA6           1 CS1.1        CS1           1 CS1.2        CS1           2 CS1.3        CS1           3 CG1.1        CG1           1 CG1.2        CG1           2 EU1.1        EU1           1 FA1.1        FA1           1 FS1.1        FS1           1 FG1.1        FG1           1 HU1.1        HU1        1, 2, 3 HU2.1        HU2           1 HU3.1        HU3           1 Page 24 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document PSL         NEI 99-01 Rev. 6 Example EAL          IC EAL HU3.2        HU3           2 HU3.3        HU3           3 HU3.4        HU3           4 HU4.1        HU4           1 HU4.2        HU4           2 HU4.3        HU4           3 HU4.4        HU4           4 HU7.1        HU7           1 HA1.1        HA1         1, 2 HA5.1        HA5           1 HA6.1        HA6           1 HA7.1        HA7           1 HS1.1        HS1           1 HS6.1        HS6           1 HS7.1        HS7           1 HG7.1        HG7           1 SU1.1        SU1           1 SU3.1        SU2           1 SU4.1        SU3           1 SU5.1        SU4        1, 2, 3 SU6.1        SU5           1 SU6.2        SU5           2 SU7.1        SU6        1, 2, 3 SU8.1        SU7         1, 2 Page 25 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document PSL         NEI 99-01 Rev. 6 Example EAL          IC EAL SA1.1        SA1          1 SA3.1        SA2          1 SA6.1        SA5          1 SA9.1        SA9          1 SS1.1        SS1          1 SS2.1        SS8          1 SS6.1        SS5          1 SG1.1        SG1          1 SG2.1        SG8          1 Page 26 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document 7.0   ATTACHMENTS 7.1   Attachment 1, Emergency Action Level Technical Bases 7.2   Attachment 2, Safe Operation & Shutdown Areas Tables R-5 & H-2 Bases Page 27 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category R - Abnormal Rad Levels / Rad Effluent EAL Group:       ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
Events of this category pertain to the following subcategories:
: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
: 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
: 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
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St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:
RU1.1                Unusual Event Reading on any Table 1[2]R-1 effluent radiation monitor > column "UE" for  60 min.
(Notes 1, 2, 3)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 29 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
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.
Page 30 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (ref. 1, 2, 3).
Escalation of the emergency classification level would be via IC RA1.
Reference(s):
: 1. C-200 Offsite Dose Calculation Manual (ODCM)
: 2. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 3. NEI 99-01 AU1 Page 31 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:
RU1.2            Unusual Event Reading on any effluent radiation monitors > 2 x the alarm setpoint established by a current radioactivity discharge permit for  60 min. (Notes 1, 2, 3)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:  If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:  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.
Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Page 32 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. Such releases are typically associated with planned batch releases from non-continuous release pathways (e.g.,
radwaste) (ref. 1).
Escalation of the emergency classification level would be via IC RA1.
Reference(s):
: 1. C-200 Offsite Dose Calculation Manual (ODCM)
: 2. NEI 99-01 AU1 Page 33 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    R - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
EAL:
RU1.3            Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate
> 2 x ODCM limits for  60 min. (Notes 1, 2)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:  If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
None Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
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 ocean water systems, etc.).
Escalation of the emergency classification level would be via IC RA1.
Page 34 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. C-200 Offsite Dose Calculation Manual (ODCM)
: 2. NEI 99-01 AU1 Page 35 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.1                Alert Reading on any Table 1[2]R-1 effluent radiation monitor > column "ALERT" for  15 min.
(Notes 1, 2, 3, 4)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Note 4         The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 36 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
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 1,000 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.
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.
Page 37 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The values of the gaseous effluent thresholds in Tables 1R-1 and 2R-1 represent 10% of the SAE values calculated in EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values (ref.
1).
Escalation of the emergency classification level would be via IC RS1.
Reference(s):
: 1. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 2. NEI 99-01 AA1 Page 38 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               1 - Radiological Effluent Initiating Condition:      Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.2          Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:  The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. NEI 99-01 AA1 Page 39 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.3             Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
This EAL is assessed per the ODCM (ref. 2)
Escalation of the emergency classification level would be via IC RS1.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. C-200 Offsite Dose Calculation Manual (ODCM)
Page 40 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 3. NEI 99-01 AA1 Page 41 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.4             Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 10 mR/hr expected to continue for  60 min.
Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.
(Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations Page 42 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 2. EPIP-10 Off-site Radiological Monitoring
: 3. NEI 99-01 AA1 Page 43 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.1                Site Area Emergency Reading on any Table 1[2]R-1 effluent radiation monitor > column "SAE" for  15 min.
(Notes 1, 2, 3, 4)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Note 4:        The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 44 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
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 1,000 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.
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.
Escalation of the emergency classification level would be via IC RG1.
Page 45 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 2. NEI 99-01 AS1 Page 46 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   R - Abnormal Rad Levels / Rad Effluent Subcategory:                1 - Radiological Effluent Initiating Condition:       Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.2           Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:   The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. NEI 99-01 AS1 Page 47 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.3             Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 100 mR/hr expected to continue for  60 min.
Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.
(Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. EPIP-10 Off-site Radiological Monitoring
: 3. NEI 99-01 AS1 Page 48 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Page 49 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.1                General Emergency Reading on any Table 1[2]R-1 effluent radiation monitor > column "GE" for  15 min.
(Notes 1, 2, 3, 4)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Note 4:        The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 50 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
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 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
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.
Reference(s):
: 1. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 2. NEI 99-01 AG1 Page 51 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               1 - Radiological Effluent Initiating Condition:      Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.2          General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:  The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. NEI 99-01 AG1 Page 52 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.3             General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 1,000 mR/hr expected to continue for  60 min.
Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.
(Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Page 53 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. EPIP-10 Off-site Radiological Monitoring
: 3. NEI 99-01 AG1 Page 54 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               R - Abnormal Rad Levels / Rad Effluent Subcategory:            2 - Irradiated Fuel Event Initiating Condition:   UNPLANNED loss of water level above irradiated fuel EAL:
RU2.1        Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following:
* SFP or reactor cavity level instrument
* SFP or reactor cavity low level alarm
* Report of dropping level in SFP or reactor cavity AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table 1[2]R-2 radiation monitors Table 1R-2 Unit 1 Refueling Area Radiation Monitors ARM #         Channel #                 Description RE26-4-1             4        Containment CIS Monitor RE26-6-1             6        Containment CIS Monitor RE26-7-1             7        Spent Fuel Pool RE26-8-1             8        Refueling Canal RE26-36-1           36        Operating Deck Page 55 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-2 Unit 2 Refueling Area Radiation Monitors ARM #      RMCS Channel                  Description RIM-26-2          GAG-002       CCW RIM-26-4          GAG-004       Containment CIS Monitor B RIM-26-6          GAG-006       Containment CIS Monitor D RIM-26-7          GAG-007       Spent Fuel Pool RIM-26-8          GAG-008       Spent Fuel Pool RIM-26-9          GAG-009       Spent Fuel Pool RIM-26-10         GAG-010       Spent Fuel Pool RIM-26-11         GAG-011       Spent Fuel Pool RIM-26-12         GAG-012       Spent Fuel Pool RIM-26-33         GAG-033       Refueling Canal Mode Applicability:
All Definition(s):
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- Refueling cavity, fuel transfer canal, and spent fuel pool (SFP), but not including the reactor vessel, comprise the refueling pathway.
Basis:
This IC addresses a decrease 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 decrease 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). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
Page 56 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The specified area radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 1, 2). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING PATHWAY level are not classifiable under this EAL.
The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase 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 Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level would be via IC RA2.
Reference(s):
: 1. 1[2]-AOP-04.01 Fuel Pool Cooling System
: 2. PSL RM EAL DBD Radiation Monitor EAL Values Used In Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 AU2 Page 57 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.1          Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC).
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.
REFUELING PATHWAY- Refueling cavity, fuel transfer canal, and spent fuel pool (SFP), but not including the reactor vessel, comprise the refueling pathway.
Basis:
This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the REFUELING PATHWAY. 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 EAL 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 EU1.
Escalation of the emergency would be based on either Category R or C EALs.
This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase 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.
Page 58 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Category C during the Cold Shutdown and Refueling modes.
Reference(s):
: 1. NEI 99-01 AA2 Page 59 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               R - Abnormal Rad Levels / Rad Effluent Subcategory:            2 - Irradiated Fuel Event Initiating Condition:   Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.2         Alert Damage to irradiated fuel resulting in a release of radioactivity AND VALID HIGH alarm on any of the following:
* Any Table 1[2]R-2 area radiation monitors
* RSC 26-4 [RSC-26-12] Fuel Handling Building Exhaust Vent Table 1R-2 Unit 1 Refueling Area Radiation Monitors ARM #         Channel #                  Description RE26-4-1             4         Containment CIS Monitor RE26-6-1             6         Containment CIS Monitor RE26-7-1             7         Spent Fuel Pool RE26-8-1             8         Refueling Canal RE26-36-1           36         Operating Deck Page 60 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-2 Unit 2 Refueling Area Radiation Monitors ARM #       RMCS Channel                 Description RIM-26-2          GAG-002       CCW RIM-26-4          GAG-004       Containment CIS Monitor B RIM-26-6          GAG-006       Containment CIS Monitor D RIM-26-7          GAG-007       Spent Fuel Pool RIM-26-8          GAG-008       Spent Fuel Pool RIM-26-9          GAG-009       Spent Fuel Pool RIM-26-10         GAG-010       Spent Fuel Pool RIM-26-11         GAG-011       Spent Fuel Pool RIM-26-12         GAG-012       Spent Fuel Pool RIM-26-33         GAG-033       Refueling Canal Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC).
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
The specified radiation monitors are those expected to see increased area radiation levels as a result of damage to irradiated fuel (ref. 1, 2).
This EAL addresses events that have caused 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 EAL 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 Page 61 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases CONFINEMENT BOUNDARY is classified in accordance with IC EU1.
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).
Escalation of the emergency would be based on either Category R or C ICs.
Reference(s):
: 1. 1[2]-AOP-04.01 Fuel Pool Cooling System
: 2. PSL RM EAL DBD Radiation Monitor EAL Values Used In Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 AA2 Page 62 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.3          Alert Lowering of spent fuel pool level to 46 ft. 3 in. [46 ft. 5 in.] (Level 2) on LI-04-1[2]A or LI 1[2]B Mode Applicability:
All Definition(s):
None Basis:
This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. 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.
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 classification level would be via ICs RS1 or RS2.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (LI-04-1[2]A and LI-04-1[2]B) capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 46 ft. 3 in. [46 ft. 5 in.]) and SFP level at the top of the fuel racks (Level 3 - 36 ft. 3 in. [36 ft. 5 in.]). PSL has designated Level 3 as 1 ft. above the highest point of any fuel rack (37 ft. 3 in. [37 ft. 5 in.]) (ref. 1).
Reference(s):
: 1. EC-280519 [280520] - SPF Level Inst. Per NRC Order EA-12-051
: 2. NEI 99-01 AA2 Page 63 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Spent fuel pool level at the top of the fuel racks EAL:
RS2.1          Site Area Emergency Lowering of spent fuel pool level to 37 ft. 3 in. [37 ft. 5 in.] (Level 3) on LI-04-1[2]A or LI 1[2]B Mode Applicability:
All Definition(s):
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.
Basis:
This EAL 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.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (LI-04-1[2]A and LI-04-1[2]B) capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 46 ft. 3 in. [46 ft. 5 in.]) and SFP level at the top of the fuel racks (Level 3 - 36 ft. 3 in. [36 ft. 5 in.]). PSL has designated Level 3 as 1 ft. above the highest point of any fuel rack (37 ft. 3 in. [37 ft. 5 in.]) (ref. 1).
Reference(s):
: 1. EC-280519 [280520] - SPF Level Inst. Per NRC Order EA-12-051
: 2. NEI 99-01 AS2 Page 64 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  2 - Irradiated Fuel Event Initiating Condition:         Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
RG2.1             General Emergency Spent fuel pool level cannot be restored to at least 37 ft. 3 in. [37 ft. 5 in.] (Level 3) on LI-04-1[2]A or LI-04-1[2]B for  60 min. (Note 1)
Note 1: The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
None Basis:
This EAL 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 EAL would likely not be met until well after another General Emergency EAL was met; however, it is included to provide classification diversity.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (LI-04-1[2]A and LI-04-1[2]B) capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 46 ft. 3 in. [46 ft. 5 in.]) and SFP level at the top of the fuel racks (Level 3 - 36 ft. 3 in. [36 ft. 5 in.]). PSL has designated Level 3 as 1 ft. above the highest point of any fuel rack (37 ft. 3 in. [37 ft. 5 in.]) (ref. 1).
Reference(s):
: 1. EC-280519 [280520] - SPF Level Inst. Per NRC Order EA-12-051
: 2. NEI 99-01 AG2 Page 65 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              3 - Area Radiation Levels Initiating Condition:     Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.1          Alert Dose rate > 15 mR/hr in EITHER of the following areas:
* Control Room
* Central Alarm Station Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
Basis:
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
Escalation of the emergency classification level would be via Category R, C or F ICs.
Reference(s):
: 1. 1[2]-AOP-26.02 Area Radiation Monitors
: 2. NEI 99-01 AA3 Page 66 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                       R - Abnormal Rad Levels / Rad Effluent Subcategory:                    3 - Area Radiation Levels Initiating Condition:           Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.2             Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-3 room or area (Note 5)
Note 5:   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.
Note 15:  The Intake Structure includes the Intake Cooling Waiter (ICW) system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table R-3 Safe Operation & Shutdown Rooms/Areas Room/Area                                        Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                          1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and                              1, 2, 3, 4, 5 Control Rooms)
Cable Spreading Room                                                      1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                                    1, 2, 3, 4, 5 Steam Trestle Area                                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                                1, 2, 3, 4, 5 Component Cooling Water Area                                              1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                      1, 2, 3, 4, 5 Refueling Water Tank                                                      1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                         1, 2, 3, 4 Containment                                                                    4, 5 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
Page 67 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Page 68 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For RA3.2, 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 at the time of the elevated radiation levels. 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, 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 increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
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 Category R, C or F ICs.
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
Reference(s):
: 1. Attachment 2 Safe Operation & Shutdown Areas Tables 1R-3 & H-2 Bases Page 69 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 2. NEI 99-01 AA3 Page 70 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature  200&#xba;F); EALs in this category are applicable only in one or more cold operating modes.
Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 -
Cold Shutdown, 6 - Refueling, DEF - Defueled).
The events of this category pertain to the following subcategories:
: 1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
: 2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV emergency buses.
: 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.
: 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
: 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
Page 71 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of multiple safety system trains warranting classification.
Page 72 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  1 - RCS Level Initiating Condition:         UNPLANNED loss of RCS inventory EAL:
CU1.1             Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
Basis:
RCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.
With the plant in Cold Shutdown, RCS water level is normally maintained within a pressurizer level control band (ref. 1). However, if RCS level is being controlled below the normal pressurizer level control band, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.
With the plant in Refueling mode, RCS water level is normally maintained at or above the reactor vessel flange (ref. 2).
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 RCS level concurrent with indications of coolant leakage. Either of these conditions are considered to be a potential degradation of the level of safety of the plant.
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.
This EAL recognizes that the minimum required RCS 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, Page 73 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
Reference(s):
: 1. 1[2]-GOP-305 Reactor Plant Cooldown - Hot Standby to Cold Shutdown
: 2. 1[2]-GOP-365 Refueling Sequencing Guidelines
: 3. NEI 99-01 CU1 Page 74 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      UNPLANNED loss of RCS inventory EAL:
CU1.2          Unusual Event RCS water level cannot be monitored AND EITHER:
* UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter changes 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.
Basis:
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 RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.
Page 75 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
This EAL addresses a condition where all means to determine RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1) (ref. 1). 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 RCS.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. NEI 99-01 CU1 Page 76 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      Significant Loss of RCS inventory EAL:
CA1.1          Alert Loss of RCS inventory as indicated by RCS water level < minimum required for continued LPSI pump operation Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
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.
For this EAL, a lowering of RCS water level below the minimum required for continued LPSI pump operation indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
The minimum RCS level for continued LPSI operation is (ref. 1):
* Tygon hose 29 ft. 9 1/2 in.
* LI-1103 Level Cold Condition 0%
* LI-1117 RX REFUELING WATER LEVEL (NR) 24 1/2 in.
* LI-1117-1 RX REFUELING WATER LEVEL (WR) 24 1/2 in.
Although related, this EAL 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 Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
If RCS water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
Reference(s):
: 1. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 2. NEI 99-01 CA1 Page 77 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 1 - RCS Level Initiating Condition:        Significant Loss of RCS inventory EAL:
CA1.2            Alert RCS water level cannot be monitored for  15 min. (Note 1)
AND EITHER
* UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
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St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
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.
For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (ref. 1). 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 RCS .
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.
If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. NEI 99-01 CA1 Page 79 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              1 - RCS Level Initiating Condition:     Loss of RCS inventory affecting core decay heat removal capability EAL:
CS1.1          Site Area Emergency 0 in. on LI-1117 or LI-1117-1 Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
Basis:
This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. 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 RCS level. If RCS level cannot be restored, fuel damage is probable.
0 in. on LI-1117 or LI-1117-1 is the lowest RCS level that can be monitored in the cold condition and corresponds to the bottom of the RCS hot leg (ref. 1).
This EAL addresses 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 ICs CG1 or RG1.
Reference(s):
: 1. PSL UFSAR Table 7.5-2 Safety Related Display Instrumentation
: 2. NEI 99-01 CS1 Page 80 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   C - Cold Shutdown / Refueling System Malfunction Subcategory:                1 - RCS Level Initiating Condition:       Loss of RCS inventory affecting core decay heat removal capability EAL:
CS1.2           Site Area Emergency RCS level cannot be monitored for  30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-1 sump or tank level of sufficient magnitude to indicate core uncovery
* Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
* Containment High Range Radiation Monitor reading > 4.8E+04 R/hr (Refueling Mode)
* Erratic source range monitor indications Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
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St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Basis:
This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. 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 level cannot be restored, fuel damage is probable.
In this EAL, 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 RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation (ref. 1). 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 RCS .
If the make-up rate to the RCS unexplainably increases above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory.
In the Refueling mode, as water level in the reactor vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in on-scale indications on area radiation monitors. A reading > 4.8E+04 R/hr on Containment High Range Radiation Monitors (CHRRMs) is indicative of RCS water level at the top of the core while in the Refueling mode (ref. 2).
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
This EAL addresses 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 CG1 or RG1 Page 82 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 CS1 Page 83 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  1 - RCS Level Initiating Condition:         Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:
CG1.1             General Emergency RCS level cannot be monitored for  30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-1 sump or tank level of sufficient magnitude to indicate core uncovery
* Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
* Containment High Range Radiation Monitor reading > 4.8E+04 R/hr (Refueling Mode)
* Erratic source range monitor indications AND Any Containment Challenge indication, Table C-2 Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 6:   If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.
Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Page 84 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table C-2 Containment Challenge Indications
* CONTAINMENT CLOSURE not established (Note 6)
* CTMT hydrogen concentration > 4%
* UNPLANNED increase in CTMT pressure Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
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.
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.
Basis:
This IC addresses the inability to restore and maintain RCS 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 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 level cannot be restored, fuel damage is probable.
The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS (ref. 1).
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St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases If the make-up rate to the RCS unexplainably increases above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory.
In the Refueling mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in on-scale indications on area radiation monitors. A reading > 4.8E+04 R/hr on Containment High Range Radiation Monitors (CHRRMs) is indicative of RCS water level at the top of the core while in the Refueling mode (ref. 2).
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
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.
Three conditions are associated with a challenge to containments capability to serve as an effective barrier to fission product release:
: 1. 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.
: 2. 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 of 4%). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). 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.
: 3. Any UNPLANNED increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential challenge of CONTAINMENT CLOSURE capability.
Page 86 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases UNPLANNED containment pressure increase indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
This EAL addresses 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.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. 1(2)-GMM-68.02 Emergency Closure of Containment Penetrations, Personnel Hatch and Equipment Hatch
: 3. NEI 99-01 CG1 Page 87 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 2 - Loss of Emergency AC Power Initiating Condition:        Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:
CU2.1            Unusual Event AC power capability, Table 1[2]C-3, to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 reduced to a single power source for  15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table 1[2]C-3 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer
* 1A[2A] Auxiliary Transformer back fed via Main Transformer (if already aligned)
* 1B[2B] Auxiliary Transformer back fed via Main Transformer (if already aligned)
Onsite
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
Page 88 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
Table 1[2]C-3 provides a list of offsite and onsite AC electrical power sources credited for this EAL. Offsite AC power sources annotated (if already aligned) require more than 15 minutes to establish and therefore are only credited if the source was already aligned at the time of AC power loss (ref. 1, 2, 3).
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.
An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed or back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.
This cold condition EAL is equivalent to the hot condition EAL SA1.1.
Reference(s):
: 1. Unit 1[2]Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. NEI 99-01 CU2 Page 89 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  2 - Loss of Emergency AC Power Initiating Condition:         Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:
CA2.1             Alert Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators, provided it can be aligned within the 15 minute classification criteria.
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.
Page 90 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Escalation of the emergency classification level would be via ICs CS1 or RS1.
This cold condition EAL is equivalent to the hot condition EAL RS1.1.
Reference(s):
: 1. Unit 1[2]Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. NEI 99-01 CU2 Page 91 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              3 - RCS Temperature Initiating Condition:     UNPLANNED increase in RCS temperature EAL:
CU3.1           Unusual Event UNPLANNED increase in RCS temperature to > 200&deg;F due to loss of decay heat removal capability Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
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.
Basis:
In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5 (ref. 1). If the RCS is intact, classification should be based on the RCS pressure increase criteria of CA3.1. RCS time to boil is calculated per 1[2]-AOP-03.02  (ref. 2).
This EAL addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant (ref. 1). If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the EC should also refer to EAL CA3.1.
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
This EAL 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.
Page 92 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases During an outage, the level in the reactor vessel will normally be maintained at or 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 increase in reactor coolant temperature depending on the time after shutdown (ref. 3).
Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
Reference(s):
: 1. Unit 1[2] Technical Specifications Table 1.2
: 2. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 3. 1[2]-GOP-365 Refueling Sequencing Guidelines
: 4. NEI 99-01 CU3 Page 93 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 3 - RCS Temperature Initiating Condition:        UNPLANNED increase in RCS temperature EAL:
CU3.2            Unusual Event Loss of all RCS temperature and RCS water level indication for  15 min. (Note 1)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6- Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
Basis:
This EAL addresses the inability to determine RCS temperature and level, and 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 EC should also refer to EAL CA3.1.
This EAL 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 CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
Reference(s):
: 1. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 2. NEI 99-01 CU3 Page 94 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Page 95 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   C - Cold Shutdown / Refueling System Malfunction Subcategory:                3 - RCS Temperature Initiating Condition:       Inability to maintain plant in cold shutdown EAL:
CA3.1            Alert UNPLANNED increase in RCS temperature to > 200&deg;F for > Table C-4 duration due to a loss of RCS cooling (Notes 1, 13)
OR UNPLANNED RCS pressure increase > 10 psig (does not apply to solid plant conditions)
Note 1:     The EC should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Note 13:    If an RCS heat removal system is in operation within the applicable Table C-4 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.
Table C-4        RCS Heat-up Duration Thresholds CONTAINMENT RCS Status                                                    Heat-up Duration CLOSURE Status Intact AND not N/A                            60 min.
reduced inventory Not intact OR                            Established                        20 min.
reduced inventory Not established                      0 min.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
Page 96 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Basis:
In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5 (ref. 1). If the RCS is intact, classification should be based on the RCS pressure increase criteria of CA3.1. RCS time to boil is calculated per 1[2]-AOP-03.02  (ref. 2).
The RCS is considered to be at reduced inventory when RCS level is less than or equal to 3 ft.
below the reactor vessel flange (ref. 3).
This EAL 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 an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresholds table also addresses an increase 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 increase without a substantial degradation in plant safety.
Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and
: 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.
Escalation of the emergency classification level would be via IC CS1 or RS1.
Reference(s):
: 1. Unit 1[2] Technical Specifications Table 1.2
: 2. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 3. 1[2]-GOP-365 Refueling Sequencing Guidelines
: 4. NEI 99-01 CA3 Page 97 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  4 - Loss of Vital DC Power Initiating Condition:         Loss of vital DC power for 15 minutes or longer EAL:
CU4.1             Unusual Event Indicated voltage < 112 VDC on required vital 125 VDC buses 1[2]A and 1[2]B for 15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis The vital (Class 1E) 125 V DC power system consists of two physically and electrically separated redundant buses - Bus 1A and Bus 1B. Each bus has a 60 cell battery bank with an eight-hour rating of 2400 amp-hours, with an operating range of 105 Volts to 140 Volts (ref. 1).
However, the current design minimum voltage is 112 Volts for the worst case safety load to have sufficient terminal voltage (ref. 2).
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 Page 98 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 CA1 or CA3, or an IC in Category R.
This cold condition EAL is equivalent to the hot condition EAL SS2.1.
Reference(s):
: 1. Unit 1[2] UFSAR Section 8.3.2 DC Power System
: 2. PSL-ENG-SEES-06-018
: 3. NEI 99-01 CU4 Page 99 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               C - Cold Shutdown / Refueling System Malfunction Subcategory:            5 - Loss of Communications Initiating Condition:   Loss of all onsite or offsite communications capabilities EAL:
CU5.1          Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRC communication methods Table C-5     Communication Methods State/
System                        Onsite                NRC Local Plant Radio System                                  X Commercial Telephone System                         X         X          X Plant Address (PA) System                           X Federal Telephone System (ENS)                                           X EMnet                                                         X Hot Ring Down (HRD) Telephone System                          X Shift Manager Cell Phones                                     X          X Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s):
None Page 100 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies 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.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the State of Florida and affected local communities.
The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
This cold condition EAL is equivalent to the hot condition EAL SU7.1.
Reference(s):
: 1. St. Lucie Plant Radiological Emergency Plan
: 2. NEI 99-01 CU5 Page 101 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 6 - Hazardous Event Affecting Safety Systems Initiating Condition:        Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:
CA6.1            Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
* Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
* Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9:   If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.
Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
Table C-6      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 Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Page 102 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Definition(s):
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 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 fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
Basis:
This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.
Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/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 Page 103 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases reliability of the SAFETY SYSTEM train.
VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
An event affecting equipment common to two or more trains of a safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under this EAL, as appropriate to the plant mode.
By affecting the functionality of multiple trains of a safety system, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under this EAL because the two-train impact criteria that underlie the EALs and bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/EC judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under this EAL, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and bases, and is warranted because the event was severe enough to affect the functionality of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.
Escalation of the emergency classification level would be via IC CS1 or RS1.
This cold condition EAL is equivalent to the hot condition EAL SA9.1.
Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. 0-AOP-117.02 Earthquake
: 3. 1[2]-AOP-100.01 Response to Fire
: 4. EP FAQ 2016-002
: 5. NEI 99-01 CA6 Page 104 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.
A Notification of Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.
The PSL ISFSI is located wholly within the PSL plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.
Page 105 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 E - ISFSI Subcategory:              1 - Confinement Boundary Initiating Condition:     Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EU1.1          Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HSM) > any of the following:
* 1,600 mrem/hr on the HSM front bird screen
* 10 mrem/hr on the outside HSM door centerline
* 10 mrem/hr on the HSM end shield wall exterior Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC).
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.
Basis:
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.
The existence of damage is determined by radiological survey. The specified EAL threshold values correspond to 2 times the NUHOMS Horizontal Storage Module (HSM) technical specification external surface dose rate limits (ref. 1). The technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Page 106 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Security-related events for ISFSIs are covered under ICs HU1 and HA1.
Reference(s):
: 1. Certificate of Compliance Appendix A NUHOMS HD System Generic Technical Specifications Section 5.4 HSM-H Dose Rate Evaluation Program
: 2. NEI 99-01 E-HU1 Page 107 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200&#xba;F); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly 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 fission product barriers are:
A. Fuel Clad Barrier (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System Barrier (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment Barrier (CTMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. Loss means the barrier no longer assures containment of radioactive materials. Potential Loss means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:
* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
Page 108 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.
* The fission product barrier thresholds specified within a scheme reflect plant-specific PSL design and operating characteristics.
* As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the containment, an interfacing system, or outside of the containment.
The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.
* At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the EC would have more assurance that there was no immediate need to escalate to a General Emergency.
Page 109 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 Fission Product Barrier Degradation Subcategory:              N/A Initiating Condition:     Any loss or any potential loss of either Fuel Clad or RCS EAL:
FA1.1          Alert Any loss or any potential loss of EITHER Fuel Clad or RCS barrier (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad 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 Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 Reference(s):
: 1. NEI 99-01 FA1 Page 110 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  Fission Product Barrier Degradation Subcategory:               N/A Initiating Condition:      Loss or potential loss of any two barriers EAL:
FS1.1          Site Area Emergency Loss or potential loss of any two barriers (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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.
Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.
At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
* One barrier loss and a second barrier loss (i.e., loss - loss)
* One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
* One barrier potential loss and a second barrier potential loss (i.e., potential loss -
potential loss)
At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the would have greater assurance that escalation to a General Emergency is less IMMINENT.
Reference(s):
: 1. NEI 99-01 FS1 Page 111 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 Fission Product Barrier Degradation Subcategory:              N/A Initiating Condition:     Loss of any two barriers and loss or potential loss of third barrier EAL:
FG1.1         General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.
At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
* Loss of Fuel Clad, RCS and Containment Barriers
* Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
* Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
* Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier Reference(s):
: 1. NEI 99-01 FG1 Page 112 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:
A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CTMT Radiation / RCS Activity D. CTMT Integrity or Bypass E. EC Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.
The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word None is entered in the cell.
Thresholds are assigned sequential numbers within each barrier column beginning with number one.
If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.
Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.
When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification.
Page 113 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FC)                                          Reactor Coolant System Barrier (RCS)                                                   Containment Barrier (CTMT)
Category                    Loss                         Potential Loss                                  Loss                             Potential Loss                                Loss                            Potential Loss
: 1. RVLMS indicates Sensors 4            1. An automatic or manual SIAS          1. UNISOLABLE RCS or SG tube          1. A leaking or RUPTURED SG is through 8 not covered                  actuation required by EITHER:            leakage > 50 gpm                      FAULTED outside of CTMT A                                                  (Note 12)
* UNISOLABLE RCS leakage            2. Uncontrolled RCS cooldown RCS or SG                    None                                                                                                     > 100&#xba;F/hr and RCS pressure                                                                   None Tube
* SG tube RUPTURE                      and temperature are to the left of Leakage                                                                                                                             the Maximum Subcooled (PT)
Curve (EOP Figure 1A or 1B)
: 1. REP CETs > 1200&deg;F                  2. REP CETs > 700&deg;F                                                             3. Applicable RCS and Core Heat                                              1. REP CETs > 1200&deg;F Removal (HR) Safety Function B                                             3. Applicable RCS and Core Heat Removal (HR) Safety Function None Status Check acceptance criteria not met for > 15 min.
None AND Inadequate                                            Status Check acceptance criteria                                                   AND                                                                         Functional Recovery not met for > 15 min.                                                           OTC not established                                                            (1[2]-EOP-15) for RCS and Core Heat Removal AND                                                                                                                                                          Heat Removal not effective OTC not established                                                                                                                                            within 15 min. (Note 1)
: 2. CTMT high range radiation                                                 2. Any CIAS [CIS] monitor                                                                                              2. CTMT high range radiation C           monitors Ch. # 58/59                                                          Ch. # 3/4/5/6 [RIM-26-3/4/5/6]                                                                                        monitors Ch. # 58/59
[RIM-26-40/41]                                                                > 1.5E+03 mR/hr                                                                                                       [RIM-26-40/41] > 2.7E+03 R/hr CTMT                                                            None                                                                             None                                  None
                 > 1.4E+02 R/hr Radiation /
RCS      3. Coolant activity > 300 Ci/gm Activity      dose equivalent I-131
: 2. CTMT isolation (CIS [CIAS]) is      3. CTMT pressure > 42 psig required AND EITHER:                         4. CTMT hydrogen concentration
                                                                                                                                                                                                                      > 4%
D
* CTMT integrity has been lost based on EC judgment           5. CTMT pressure > 10 psig [5.4] with CTMT                      None                                   None                                   None                                     None                                                             < one full train of containment heat Integrity or
* UNISOLABLE pathway from            removal systems operating per Bypass                                                                                                                                                                          CTMT atmosphere to the             design for  15 min. (Notes 1, 11) environment exists
: 3. Indications of RCS leakage outside of CTMT E       4. Any condition in the opinion of the EC that indicates loss of the
: 4. Any condition in the opinion of the EC that indicates potential
: 3. Any condition in the opinion of the EC that indicates loss of the RCS
: 4. Any condition in the opinion of the EC that indicates potential loss of
: 4. Any condition in the opinion of the EC that indicates loss of the
: 6. Any condition in the opinion of the EC that indicates potential loss of EC          fuel clad barrier                     loss of the fuel clad barrier           barrier                                the RCS barrier                        CTMT barrier                            the CTMT barrier Judgment Page 114 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:
None Page 115 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                      Fuel Clad Category:                     A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
: 1. RVLMS indicates Sensors 4 through 8 not covered (Note 12)
Note 12: This threshold is only applicable if all RVLMS Sensors 4 through 8 are not covered inclusive.
Definition(s):
None Basis:
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
PSL includes in its Inadequate Core Cooling instrumentation a reactor vessel level monitoring system (RVLMS) that is displayed to the operators and can measure discrete reactor vessel water levels from the top of the reactor vessel head to the top of the fuel alignment plate (Sensor #8). The bottom of this instrument's span in the reactor vessel plenum is the lowest available reactor vessel level indicator and is used in this EAL to represent approximately the top of active fuel (ref. 1).
Reference(s):
: 1. Unit 1 UFSAR Section 7.5.4 Safety Assessment System/Emergency Response Data Acquisition and Display System and Qualified Safety Parameter Display System
: 2. Unit 2 UFSAR Section 7.5.4 Instrumentation for Detection of Inadequate Core Cooling
: 3. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A Page 116 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
: 1. REP CETs > 1200&deg;F Definition(s):
None Basis:
This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 1).
Reference(s):
: 1. EPIP-11 Core Damage Assessment
: 2. NEI 99-01 Inadequate Heat Removal Loss 2.A Page 117 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Fuel Clad Category:                  B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 2. REP CETs > 700&deg;F Definition(s):
None Basis:
This reading indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
700 &deg;F is the threshold for indicating a loss of subcooling and further temperature increases will potentially result in damage fuel cladding (ref. 1).
Reference(s):
: 1. EPIP-11 Core Damage Assessment
: 2. NEI 99-01 Inadequate Heat Removal Loss 2.A Page 118 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 3. Both SG wide range level < 15% for  15 min.
AND OTC not established Definition(s):
None Basis:
There are three defined and potentially applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria conditions (ref. 1):
* HR SG Heat Sink Without SIAS
* HR SG Heat Sink With SIAS
* HR Once-Through-Cooling Wide range level in both steam generators less than 15% for  15 minutes without Once-Through Core Cooling (OTC) established indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink) (ref. 1, 2). The 15-minute criteria provides the appropriate time for establishment of the heat removal safety function and aligns with the required Functional Recovery safety function status check frequency (ref. 2). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS Barrier Potential Loss threshold B.3; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
Reference(s):
: 1. 1[2]-EOP-15 Functional Recovery
: 2. 1[2]-EOP-06 Total Loss of Feedwater TLOF
: 3. NEI 99-01 Inadequate Heat Removal Fuel Clad Potential Loss 2.B Page 119 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                C. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:
: 2. CTMT high range radiation monitors Ch. # 58/59 [RIM-26-40/41] > 1.4E+02 R/hr Definition(s):
None Basis:
Containment radiation monitor readings greater than the value shown indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of ~1% clad failure into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within Technical Specifications and are therefore indicative of fuel damage (ref. 1).
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.2 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 ECL to a Site Area Emergency.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Reference(s):
: 1. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 2. NEI 99-01 CTMT Radiation / RCS Activity FC Loss 3.A Page 120 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Fuel Clad Category:                 C. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:
: 3. Coolant activity > 300 Ci/gm dose equivalent I-131 Definition(s):
None Basis:
This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Reference(s):
: 1. NEI 99-01 CTMT Radiation / RCS Activity FC Loss 3.B Page 121 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            C. CTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 122 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
None Page 123 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Page 124 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:               Fuel Clad Category:              E. EC Judgment Degradation Threat: Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates loss of the Fuel Clad barrier Definition(s):
None Basis:
This threshold addresses any other factors that are to be used by the EC in determining whether the Fuel Clad barrier is lost.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 125 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                E. EC Judgment Degradation Threat: Potential Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates potential loss of the Fuel Clad barrier Definition(s):
None Basis:
This threshold addresses any other factors that are to be used by the EC in determining whether the Fuel Clad barrier is potentially lost. The EC should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 126 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Reactor Coolant System Category:                 A. RCS or S/G Tube Leakage Degradation Threat: Loss Threshold:
: 1. An automatic or manual SIAS actuation required by EITHER:
* UNISOLABLE RCS leakage
* SG tube RUPTURE Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold A.1 will also be met.
Reference(s):
: 1. 1[2]-EOP-03 Loss of Coolant Accident (LOCA)
: 2. 1[2]-EOP-04 Steam Generator Tube Rupture SGTR
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 127 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Reactor Coolant System Category:                 A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:
: 1. UNISOLABLE RCS or SG tube leakage > 50 gpm Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging pump, but an SIAS actuation has not occurred. The threshold is met when RCS leakage is determined to exceed 50 gpm excluding normal reductions in RCS inventory such as letdown and RCP seal leakoff (ref. 1, 2).
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If the leaking steam generator (> 50 gpm) is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold A.1 will also be met.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. 1[2]-AOP-08.02 Steam Generator Tube Leak
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 128 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Reactor Coolant System Category:                A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:
: 2. Uncontrolled RCS cooldown > 100&#xba;F/hr and RCS pressure and temperature are to the left of the Maximum Subcooled (PT) Curve (EOP Figure 1A or 1B)
Definition(s):
None Basis:
This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
The maximum allowed RCS cooldown rate per Technical Specifications is 100&#xba;F/hr with cold leg temperature > 180&#xba;F [105 &#xba;F] (ref. 1).
A RCS cooldown greater than 100&deg;F/hr below 500&deg;F is the temperature that requires implementation of Pressurized Thermal Shock (PTS) guidance (ref. 2, 3).
Reference(s):
: 1. Technical Specifications Section 3/4.4.9 Pressure/Temperature Limits
: 2. 1[2]-EOP-05 Excess Steam Demand Event
: 3. 1[2]-EOP-99 Appendices/Figures/Tables/Data Sheets, Figure 1A & 1B
: 4. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B Page 129 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
None Page 130 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Reactor Coolant System Category:                 B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 3. Both SG wide range level < 15% for  15 min.
AND OTC not established Definition(s):
None Basis:
There are three defined and potentially applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria conditions (ref. 1):
* HR SG Heat Sink Without SIAS
* HR SG Heat Sink With SIAS
* HR Once-Through-Cooling Wide range level in both steam generators less than 15% for  15 minutes without Once-Through Core Cooling (OTC) established indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink) (ref. 1, 2). The 15-minute criteria provides the appropriate time for establishment of the heat removal safety function and aligns with the required Functional Recovery safety function status check frequency (ref. 2). This condition represents a potential loss of the RCS Barrier.
In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 4B.3; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
Reference(s):
: 1. 1[2]-EOP-15 Functional Recovery
: 2. 1[2]-EOP-06 Total Loss of Feedwater TLOF
: 3. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.A Page 131 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Reactor Coolant System Category:                C. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:
: 2. Any CIAS [CIS] monitor Ch. # 3/4/5/6 [RIM-26-3/4/5/6] > 1.5E+03 mR/hr Definition(s):
None Basis:
A reading > the value shown is indicative of a breach in the RCS barrier (ref. 1).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits (0.1% failed fuel). This value is lower than that specified for Fuel Clad barrier loss threshold C.2 since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Reference(s):
: 1. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A Page 132 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            C. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 133 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
None Page 134 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Page 135 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Reactor Coolant System Category:                E. EC Judgment Degradation Threat: Loss Threshold:
: 3. Any condition in the opinion of the EC that indicates loss of the RCS barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the EC in determining whether the RCS barrier is lost.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 136 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Reactor Coolant System Category:                  E. EC Judgment Degradation Threat: Potential Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates potential loss of the RCS barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the EC in determining whether the RCS barrier is potentially lost. The EC should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 137 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Containment Category:                  A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:
: 1. A leaking or RUPTURED SG is FAULTED outside of CTMT Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss A.1and Loss A.1, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).
With the exception of the initial cooldown to < 510 &#xba;F per EOP-04, this threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump (ref. 1). These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Page 138 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate                       Yes                     No Less than or equal to 25 gpm           No classification       No classification Greater than 25 gpm              Unusual Event per SU5.1   Unusual Event per SU5.1 Requires operation of a standby Site Area Emergency per charging (makeup) pump (RCS                                    Alert per FA1.1 FS1.1 Barrier Potential Loss)
Requires an automatic or manual Site Area Emergency per ECCS (SIAS) actuation (RCS                                     Alert per FA1.1 FS1.1 Barrier Loss)
There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
Reference(s):
: 1. 1[2]-EOP-04 Steam Generator Tube Rupture SGTR
: 2. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A Page 139 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Containment Category:            A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
None Page 140 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Containment Category:            B. Inadequate Heat Removal Degradation Threat: Loss None Page 141 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                      Containment Category:                     B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 1. REP CETs > 1200&deg;F AND Functional Recovery (1[2]-EOP-15) for RCS and Core Heat Removal not effective within 15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Definition(s):
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.
Basis:
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 1).
This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered effective if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The EC should escalate the emergency classification level to a General Emergency as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
Reference(s):
: 1. EPIP-11 Core Damage Assessment
: 2. 1[2]-EOP-15 Functional Recovery Page 142 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A Page 143 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Containment Category:            C. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold:
None Page 144 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Containment Category:                  C. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:
: 2. CTMT high range radiation monitors Ch. # 58/59 [RIM-26-40/41] > 2.7E+03 R/hr Definition(s):
None Basis:
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the 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 (ref. 1).
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.
Reference(s):
: 1. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A Page 145 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Containment Category:                  D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
: 2. CTMT isolation (CIS [CIAS]) is required AND EITHER:
* CTMT integrity has been lost based on EC judgment
* UNISOLABLE pathway from CTMT atmosphere to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.
These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds. The threshold specifies (CIS [CIAS]) as the site-specific isolation signals that initiate required containment isolations to preclude fission product release from the containment (ref. 1).
First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the EC will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g.,
containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this Page 146 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.
Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term environment includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then the second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
Reference(s):
: 1. UFSAR Section 6.2.4 Containment Isolation System
: 2. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A Page 147 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Containment Category:                D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
: 3. Indications of RCS leakage outside of CTMT Definition(s):
None Basis:
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A.1 to be met.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Loss Threshold A.1.
Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.
should be sufficient to determine if RCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold Containment Loss D.2 to be met as well.
Reference(s):
: 1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.B Page 148 of 231

St Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Figure 1: Containment Integrity or Bypass Examples
                                                                                                . ::: : .. 2na Threshold~ . ::
                                                                                               <::::::   Airborne release :: ....
: : : .. f:~~ _P~!~V;'~Y.. :: :/.'.
Inside CTMT Auxiliary Building I                ;
i      Area J __ Monitor _ :
                                                                    ,re-._,,,.-,-...
Open valve Penetration     Damper~~~
                                                    ,1f t Closed Cooling Page 149 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Containment Category:                D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
: 3. CTMT pressure > 42 psig Definition(s):
None Basis:
If containment pressure exceeds the design pressure of 42 psig (ref. 1, 2), there exists a potential to lose the containment barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Reference(s):
: 1. PSL-ENG-SEIS-01-046 Unit 1 Emergency Operating Procedure Setpoint Basis Document
: 2. PSL-ENG-SEIS-01-047 Unit 2 Emergency Operating Procedure Setpoint Basis Document
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A Page 150 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Containment Category:                 D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
: 4. CTMT hydrogen concentration > 4%
Definition(s):
None Basis:
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the containment barrier.
A containment hydrogen concentration of 4% conservatively represents the lowest threshold for flammability in the presence of oxygen (ref. 1).
Reference(s):
: 1. PSL-ENG-SEIS-01-046 Unit 1 Emergency Operating Procedure Setpoint Basis Document
: 2. PSL-ENG-SEIS-01-047 Unit 2 Emergency Operating Procedure Setpoint Basis Document
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B Page 151 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                      Containment Category:                     D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
: 5. CTMT pressure > 10 psig [5.4 psig] with < one full train of CTMT heat removal systems operating per design for  15 min. (Notes 1, 11)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 11: One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700] gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550
[2,700] gpm Definition(s):
None Basis:
This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design (ref. 1, 2). The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays but not including containment venting strategies) are either lost or performing in a degraded manner.
One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700]
gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550 [2,700] gpm Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Section 7.3 Engineered Safety Features Actuation System
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C Page 152 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                Containment Category:               E. EC Judgment Degradation Threat: Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates loss of the CTMT barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the EC in determining whether the containment barrier is lost.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A Page 153 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Containment Category:                E. EC Judgment Degradation Threat: Potential Loss Threshold:
: 6. Any condition in the opinion of the EC that indicates potential loss of the CTMT barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the EC in determining whether the containment barrier is potentially lost. The EC should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A Page 154 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
: 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
: 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
: 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
: 4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown.
: 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.
: 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
: 7. EC Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the EC the latitude to classify emergency conditions consistent with the established classification criteria based upon EC judgment.
Page 155 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HU1.1           Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by PSL Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA (OCA) - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
Page 156 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
Basis:
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 &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
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.
The first threshold references the Security Shift Supervision 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 &sect; 2.39 information.
The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with SFI-2416 Security and Nuclear Operations Safeguards Contingencies (ref. 1).
The third threshold 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 SY-AA-102-1014 Threat Assessment and Reporting (ref. 2).
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security and Nuclear Operations Safeguards Contingencies (ref. 1).
Page 157 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC HA1.
Reference(s):
: 1. SFI-2416 Security and Nuclear Operations Safeguards Contingencies
: 2. SY-AA-102-1014 Threat Assessment and Reporting
: 3. NEI 99-01 HU1 Page 158 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HA1.1          Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by PSL Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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 - One or more 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.
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
Page 159 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
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 (OROs), allowing them to be better prepared should it be necessary to consider further actions.
This EAL 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 &sect; 73.71 or 10 CFR &sect; 50.72.
The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA.
The second threshold 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 SY-AA-102-1014 Threat Assessment and Reporting (ref. 2).
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.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security and Nuclear Operations Safeguards Contingencies (ref. 1).
Escalation of the emergency classification level would be via IC HS1.
Page 160 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. SFI-2416 Security and Nuclear Operations Safeguards Contingencies
: 2. SY-AA-102-1014 Threat Assessment and Reporting
: 3. NEI 99-01 HA1 Page 161 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the PROTECTED AREA EAL:
HS1.1          Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by PSL Security Shift Supervision Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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 - One or more 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.
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
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 (ref. 1, 2).
Page 162 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
This EAL 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 &sect; 73.71 or 10 CFR &sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the SFI-2416 Security and Nuclear Operations Safeguards Contingencies (ref. 1).
Reference(s):
: 1. SFI-2416 Security and Nuclear Operations Safeguards Contingencies
: 2. SY-AA-102-1014 Threat Assessment and Reporting
: 3. NEI 99-01 HS1 Page 163 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              2 - Seismic Event Initiating Condition:     Seismic event greater than OBE levels EAL:
HU2.1           Unusual Event Seismic event > OBE as indicated by S-36[S-44] CNTMT SEISMIC MONITOR HIGH LIMIT alarm Mode Applicability:
All Definition(s):
None 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). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) 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., lateral accelerations in excess of 0.05g). The Shift Manager or EC may seek external verification if deemed appropriate (e.g., a call to the U.S. Geological Survey (USGS), check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
For both Unit 1 and Unit 2, the OBE ground acceleration threshold is > 0.05g horizontal and vertical (ref. 3). The Control Rooms have real time OBE exceedance alarm indications (S-36[S-44] CNTMT SEISMIC MONITOR HIGH LIMIT alarm) (ref. 1, 2, 3, 4).
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Reference(s):
: 1. 0-AOP-117.02 Earthquake
: 2. 1[2] S-36[S-44] CNTMT SEISMIC MONITOR HIGH LIMIT
: 3. Unit 1 USFSAR Section 2.5.2.11 Operating Basis Earthquake
: 4. Unit 2 USFSAR Section 2.5.2.7 Operating Basis Earthquake
: 4. NEI 99-01 HU2 Page 164 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.1          Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a tornado striking (touching down) within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under IC CA6 or SA9.
A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.
Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. NEI 99-01 HU3 Page 165 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.2           Unusual Event 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 Mode Applicability:
All Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL 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 (ref. 1, 2).
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
Refer to EAL CA6.1 or SA9.1 for internal flooding affecting more than one SAFETY SYSTEM train.
Reference(s):
Page 166 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. 1[2]-AOP-24.01 RAB Flooding
: 2. NEI 99-01 HU3 Page 167 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.3          Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
Reference(s):
: 1. NEI 99-01 HU3 Page 168 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.4          Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)
Note 7:  This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Mode Applicability:
All Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL 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, S or C.
Reference(s):
: 1. NEI 99-01 HU3 Page 169 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                  4 - Fire Initiating Condition:         FIRE potentially degrading the level of safety of the plant EAL:
HU4.1             Unusual Event A FIRE is not extinguished within 15 min. of any of the following fire detection indications (Note 1):
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire zone alarms or indications
* Field verification of a single fire zone alarm AND The FIRE is located within any Table H-1 area Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table H-1 PSL Fire Areas
* Reactor Containment Building/Shield Building
* Control Room
* Reactor Auxiliary Building (RAB)
* Diesel Generator Building
* Component Cooling Water (CCW)
* Intake Structure
* Intake Cooling Water (ICW)
* Steam Trestle Area
* Yard Areas:
o RWT o CST o DOST Mode Applicability:
All Definition(s):
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.
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the Page 170 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
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.
The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system zone alarms or field validation of a single fire zone alarm. The zone alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.
Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).
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 alarm, indication or report.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. 1[2]-AOP-AOP-100.11 Response to Fire - Fire Zones
: 3. NEI 99-01 HU4 Page 171 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                 4 - Fire Initiating Condition:        FIRE potentially degrading the level of safety of the plant EAL:
HU4.2            Unusual Event Receipt of a single fire zone alarm (i.e., no other indications of a FIRE)
AND The fire zone alarm is indicating a FIRE within any Table H-1 area (excluding containment)
(Note 14)
AND The existence of a FIRE is not verified (proved or disproved) within 30 min. of alarm receipt (Note 1)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 14: A containment fire zone alarm is considered VALID upon receipt of multiple zones (more than one) actuated on CP-376 panel.
Table H-1 PSL Fire Areas
* Reactor Containment Building/Shield Building
* Control Room
* Reactor Auxiliary Building (RAB)
* Diesel Generator Building
* Component Cooling Water (CCW)
* Intake Structure
* Intake Cooling Water (ICW)
* Steam Trestle Area
* Yard Areas:
o RWT o CST o DOST Mode Applicability:
All Definition(s):
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.
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct Page 172 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
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.
The 30-minute requirement begins upon receipt of a single VALID fire detection system zone alarm (excluding containment). The zone alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30-minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15-minute requirement beginning with the verification of the fire by field report.
Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).
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.
With regard to containment fire zone alarms, there is constant air movement in containment due to the operation of the CAH system. The operating cooling units are drawing air to the units past the smoke detectors. It can be reasonably expected that a fire that burns for 15 minutes would produce sufficient products of combustion to cause fire zone detectors in multiple zones to alarm. Therefore a single containment fire zone alarm is not considered VALID.
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 HU4.1 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.
Basis-Related Requirements from Appendix R (Note; While PSL is an NFP-805 plant, the following excerpt from 10CFR 50 Appendix R provides the bases for the acceptability of the use of a 30-minute validation time for a single fire zone alarm.)
Appendix R to 10 CFR 50, states in part:
Page 173 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify (prove or disprove) 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 CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. 1[2]-AOP-AOP-100.11 Response to Fire - Fire Zones
: 3. NEI 99-01 HU4 Page 174 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                 4 - Fire Initiating Condition:        FIRE potentially degrading the level of safety of the plant EAL:
HU4.3            Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
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 - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
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.
In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. NEI 99-01 HU4 Page 175 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               4 - Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.4           Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:
All Definition(s):
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 - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
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.
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.
The Shift Fire Brigade Advisor or Shift Fire Brigade Leader will assess whether the fire conditions warrant outside assistance (ref. 1).
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. NEI 99-01 HU4 Page 176 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                      H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                   5 - Hazardous Gases Initiating Condition:          Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
HA5.1            Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5)
Note 5:  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.
Note 15: The Intake Structure includes the Intake Cooling Waiter (ICW) system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area                                        Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                          1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and                              1, 2, 3, 4, 5 Control Rooms)
Cable Spreading Room                                                      1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                                    1, 2, 3, 4, 5 Steam Trestle Area                                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                                1, 2, 3, 4, 5 Component Cooling Water Area                                              1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                      1, 2, 3, 4, 5 Refueling Water Tank                                                      1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                         1, 2, 3, 4 Containment                                                                    4, 5 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown Page 177 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
Basis:
This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.
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 at the time of the gaseous release. 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 ECs 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.
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
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 Page 178 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases lead to breathing difficulties, unconsciousness or even death. Steam can be an asphyxiant if present in sufficient concentration to reduce oxygen levels below normal.
This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
Reference(s):
: 1. Attachment 2 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases
: 2. NEI 99-01 HA5 Page 179 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                H - Hazards and Other Conditions Affecting Plant Safety Subcategory:             6 - Control Room Evacuation Initiating Condition:    Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA6.1          Alert An event has resulted in the Control Room being evacuated Mode Applicability:
All Definition(s):
None 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.
The Control Room is deemed to have been evacuated when the last licensed operator leaves the Control Room.
Escalation of the emergency classification level would be via IC HS6.
Reference(s):
: 1. 1[2]-ONP-100.02 Control Room Inaccessibility
: 2. NEI 99-01 HA6 Page 180 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                  6 - Control Room Evacuation Initiating Condition:         Inability to control a key safety function from outside the Control Room EAL:
HS6.1             Site Area Emergency An event has resulted in the Control Room being evacuated AND Control of any of the following key safety functions is not re-established within 15 min. of the last licensed operator leaving the Control Room (Note 1):
Reactivity (modes 1, 2 and 3 only)
Core cooling RCS heat removal Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.
The determination of whether or not control is established at the remote safe shutdown location(s) is based on EC judgment. The EC is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
The Control Room is deemed to have been evacuated when the last licensed operator leaves the Control Room.
Establishment of the reactivity safety function is only applicable in Modes 1, 2 and 3. Sufficient shutdown margin has already been established once in modes 4, 5 and 6 (ref. 2).
Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):
Page 181 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 1. 1[2]-ONP-100.02 Control Room Inaccessibility
: 2. NRC EP FAQ 2015-014
: 3. NEI 99-01 HS6 Page 182 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               7 - EC Judgment Initiating Condition:      Other conditions existing that in the judgment of the EC warrant declaration of a Unusual Event EAL:
HU7.1           Unusual Event Other conditions exist which, in the judgment of the EC, 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.
Mode Applicability:
All Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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 EC to fall under the emergency classification level description for an Unusual Event.
Reference(s):
: 1. NEI 99-01 HU7 Page 183 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              7 - EC Judgment Initiating Condition:     Other conditions exist that in the judgment of the EC warrant declaration of an Alert EAL:
HA7.1          Alert Other conditions exist which, in the judgment of the EC, 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.
Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
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 EC to fall under the emergency classification level description for an Alert.
Page 184 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Page 185 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. NEI 99-01 HA7 Page 186 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               7 - EC Judgment Initiating Condition:      Other conditions existing that in the judgment of the EC warrant declaration of a Site Area Emergency EAL:
HS7.1           Site Area Emergency Other conditions exist which, in the judgment of the EC, 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 Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 EC to fall under the emergency classification level description for a SITE AREA EMERGENCY.
Page 187 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. NEI 99-01 HS7 Page 188 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              7 - EC Judgment Initiating Condition:     Other conditions exist that in the judgment of the EC warrant declaration of a General Emergency EAL:
HG7.1                 General Emergency Other conditions exist which, in the judgment of the EC, 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.
Mode Applicability:
All Definition(s):
HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
Page 189 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the EC to fall under the emergency classification level description for a GENERAL EMERGENCY.
Reference(s):
: 1. NEI 99-01 HG7 Page 190 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200&#xba;F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.
The events of this category pertain to the following subcategories:
: 1. Loss of Emergency AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV emergency buses.
: 2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
: 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.
: 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (~1% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
: 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.
Page 191 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to properly result in reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
: 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
: 8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.
: 9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system train performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.
Page 192 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 1 - Loss of Emergency AC Power Initiating Condition:        Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:
SU1.1            Unusual Event Loss of all offsite AC power capability, Table 1[2]S-1, to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. SU3.1 Table 1[2]S-1 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer Onsite
* 1A[2A] Auxiliary Transformer
* 1B[2B] Auxiliary Transformer
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Table 1[2]S-1 provides a list of offsite and onsite AC electrical power sources credited for this EAL. The onsite AC power source annotated (if already aligned) requires more than 15 minutes to establish and therefore is only credited if the source was already aligned at the time of AC power loss (ref. 1, 2, 3).
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.
Page 193 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases For emergency classification purposes, capability means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level would be via IC SA1.
Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. NEI 99-01 SU1 Page 194 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  1 - Loss of Emergency AC Power Initiating Condition:         Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:
SA1.1             Alert AC power capability, Table 1[2]S-1, to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 reduced to a single power source for  15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table 1[2]S-1 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer Onsite
* 1A[2A] Auxiliary Transformer
* 1B[2B] Auxiliary Transformer
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 195 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
Table 1[2]S-1 provides a list of offsite and onsite AC electrical power sources credited for this EAL. The onsite AC power source annotated (if already aligned) requires more than 15 minutes to establish and therefore is only credited if the source was already aligned at the time of AC power loss (ref. 1, 2, 3).
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 SU1.
An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being cross tied to the other unit .
A loss of emergency power sources (e.g., onsite diesel generators and unit cross ties) with a single train of emergency buses being fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level would be via IC SS1.
This hot condition EAL is equivalent to the cold condition EAL CU2.1.
Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. NEI 99-01 SA1 Page 196 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   S - System Malfunction Subcategory:                1 - Loss of Emergency AC Power Initiating Condition:       Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:
SS1.1           Site Area Emergency Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
Note 1: The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators, provided it can be aligned within the 15 minute classification criteria (ref. 4, 5).
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.
This hot condition EAL is equivalent to the cold condition EAL CA2.1.
Page 197 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. 1[2]-FSG-05 FLEX Implementation
: 6. NEI 99-01 SS1 Page 198 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  S -System Malfunction Subcategory:               1 - Loss of Vital AC Power Initiating Condition:      Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:
SG1.1           General Emergency Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 AND REP CETs > 1200&#xba;F Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators (ref. 4, 5).
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 6).
This IC addresses a prolonged loss of all power sources to AC emergency buses that results in degraded core cooling. 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 eventually lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
Page 199 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases For extended loss of emergency bus AC power events that do not result in a breach of the RCS barrier, this 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.
The EAL will 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.
Reference(s):
: 1. Unit 1[2]Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. 1[2]-FSG-05 FLEX Implementation
: 6. EPIP-11 Core Damage Assessment
: 7. NEI 99-01 SG1 Page 200 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  2 - Loss of Vital DC Power Initiating Condition:         Loss of all vital DC power for 15 minutes or longer EAL:
SS2.1             Site Area Emergency Indicated voltage < 112 VDC on both vital 125 VDC buses 1[2]A and 1[2]B for  15 min.
(Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
The vital (Class 1E) 125 V DC power system consists of two physically and electrically separated redundant buses - Bus 1A[2A] and Bus 1B[2B]. Each bus has a 60 cell battery bank with an eight-hour rating of 2400 amp-hours, with an operating range of 105 Volts to 140 Volts (ref. 1). However, the current design minimum voltage is 112 Volts for the worst case safety load to have sufficient terminal voltage (ref. 2).
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 SG2.
This hot condition EAL equivalent of the cold condition EAL CU4.1.
Reference(s):
Page 201 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 1. Unit 1[2] UFSAR Section 8.3.2 DC Power System
: 2. PSL-ENG-SEES-06-018
: 3. NEI 99-01 SS8 Page 202 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S -System Malfunction Subcategory:                  2 - Loss of Vital DC Power Initiating Condition:         Loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL:
SG2.1             General Emergency Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
AND Indicated voltage is < 112 VDC on both vital 125 VDC buses 1[2]A and 1[2]B for 15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
This IC addresses a concurrent and prolonged loss of both emergency AC and vital DC power.
A loss of all emergency 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 emergency AC and vital DC power will lead to multiple challenges to fission product barriers.
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators (ref. 4, 5).
Page 203 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The Vital (Class 1E) 125 V DC power system consists of two physically and electrically separated redundant buses - Bus 1[2]A and Bus 1[2]B. Each bus has a 60 cell battery bank with an eight-hour rating of 2300 amp-hours, with an operating range of 105 Volts to 140 Volts (ref. 6). However, the current design minimum voltage is 112 Volts for the worst case safety load to have sufficient terminal voltage (ref. 7).
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 both EAL thresholds are met.
Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. 1[2]-FSG-05 FLEX Implementation
: 6. Unit 1[2] UFSAR Section 8.3.2 DC Power System
: 7. PSL-ENG-SEES-06-018
: 8. NEI 99-01 SG8 Page 204 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  3 - Loss of Control Room Indications Initiating Condition:         UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1             Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table S-2    Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CET temperature
* Level in at least one SG
* Auxiliary feedwater flow to at least one SG Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
Page 205 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
Applicable safety system parameters are listed in Table S-2.
The Safety Assessment System (SAS) and Emergency Response Data Acquisition and Display System (ERDADS) serve as redundant indicators which may be utilized as compensatory measures in lieu of the Control Room indicators associated with safety functions (ref. 1, 2).
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 one or more 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.
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 one or more 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 one or more 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 RCS 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 IC SA3.
Reference(s):
: 1. Unit 1[2] UFSAR Section 7.5 Safety Related Display Instrumentation
: 2. Unit 2 UFSAR Appendix 7.5A Safety Assessment System/Emergency Response Data Acquisition and Display System
: 3. NEI 99-01 SU2 Page 206 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 3 - Loss of Control Room Indications Initiating Condition:        UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1            Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for  15 min. (Note 1)
AND Any significant transient is in progress, Table S-3 Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table S-2     Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CET temperature
* Level in at least one SG
* Auxiliary feedwater flow to at least one SG Table S-3    Significant Transients
* Electrical load rejection > 25% full electrical load
* Reactor Trip
* SIAS actuation Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Page 207 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
Basis:
Applicable safety system parameters are listed in Table S-2.
Significant transients are listed in Table S-3.
The Safety Assessment System (SAS) and Emergency Response Data Acquisition and Display System (ERDADS) serve as redundant indicators which may be utilized as compensatory measures in lieu of the Control Room indicators associated with safety functions (ref. 1, 2).
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 one or more 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.
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 Page 208 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases determine one or more 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 one or more 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 RCS 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 RS1 Reference(s):
: 1. Unit 1[2] UFSAR Section 7.5 Safety Related Display Instrumentation
: 2. Unit 2 UFSAR Appendix 7.5A Safety Assessment System/Emergency Response Data Acquisition and Display System
: 3. NEI 99-01 SA2 Page 209 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               S - System Malfunction Subcategory:            4 - RCS Activity Initiating Condition:   Reactor coolant activity greater than Technical Specification allowable limits EAL:
SU4.1         Unusual Event Sample analysis indicates that a reactor coolant activity value is > any of the following allowable limits:
* 60 Ci/gm dose equivalent I-131 instantaneous
* 1.0 Ci/gm dose equivalent I-131 for > 48 hours
* 518.9 Ci/gm dose equivalent Xe-133 for > 48 hours Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications (ref. 1, 2, 3). This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.
Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.
Reference(s):
: 1. Technical Specification 3.4.8 RCS Specific Activity
: 2. 1[2]-AOP-01.06 Excessive RCS Activity
: 3. CY-SL-108-005 High Activity in Reactor Coolant System
: 4. NEI 99-01 SU3 Page 210 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  5 - RCS Leakage Initiating Condition:         RCS leakage for 15 minutes or longer EAL:
SU5.1             Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for  15 min.
OR RCS identified leakage > 25 gpm for  15 min.
OR Leakage from the RCS to a location outside containment > 25 gpm for  15 min.
(Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
Once the RCS leak rate has been quantified to be greater than the specified value, failure to isolate the leak within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the time of leak rate quantification, 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.
The first and second EAL conditions 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). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.
The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition 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 a relief valve does not warrant an emergency classification. An emergency classification would be required Page 211 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level would be via ICs of Recognition Category R or F.
Reference(s):
: 1. Technical Specification Definitions Section 1.0
: 2. Technical Specification 3.4.6 RCS Leakage
: 3. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. NEI 99-01 SU4 Page 212 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  6 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor EAL:
SU6.1            Unusual Event An automatic trip did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip (RX Trip Pushbuttons) are successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria met (Note 8)
Note 8:   A manual trip 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.
Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
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.
Basis:
This EAL addresses a failure of the RPS to initiate or complete an automatic trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip by pushing the RX Trip Pushbuttons). If this manual action is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.
A manual action at the reactor control console (RTGB) is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip by pushing the RX Trip Pushbuttons). This action does not include opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally, 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.
Page 213 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The plant response to the failure of an automatic trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA5.
Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5 or FA1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to shutdown the reactor, the event escalates to the Alert under EAL SA6.1.
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. NEI 99-01 SU5 Page 214 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                      S - System Malfunction Subcategory:                   6 - RPS Failure Initiating Condition:          Automatic or manual trip fails to shut down the reactor EAL:
SU6.2             Unusual Event A manual trip (RX Trip Pushbuttons) did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met (Note 8)
AND A subsequent automatic trip or manual trip action (RX Trip Pushbuttons) is successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria met Note 8:    A manual trip 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.
Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
None Basis:
This EAL addresses a failure of a manual reactor trip to shutdown the reactor, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual actions at other location(s) away from the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip by opening CEDM MG set feeder breakers or opening the reactor trip feeder breakers locally) (ref. 3). However, those actions are not credited as a successful manual reactor trip for this EAL.
Depending upon several factors, the initial effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.
A manual action at the reactor control console (RTGB) is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip by pushing the RX Trip Pushbuttons). This action does not include opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally, manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or Page 215 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles.
The plant response to the failure of a manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5 or FA1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. NEI 99-01 SU5 Page 216 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  2 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:
SA6.1            Alert An automatic or manual trip (RX Trip Pushbuttons) did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met AND Subsequent automatic or manual trip actions (RX Trip Pushbuttons) are not successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met (Note 8)
Note 8:   A manual trip 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.
Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
None Basis:
This EAL addresses a failure of the RPS to initiate or complete an automatic reactor trip or failure of a manual reactor trip to shutdown the reactor, 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 console (RTGB) is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip by pushing the RX Trip Pushbuttons). This action does not include opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally, manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control console.
Page 217 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.
It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. NEI 99-01 SA5 Page 218 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 S - System Malfunction Subcategory:              2 - RPS Failure Initiating Condition:     Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL:
SS6.1           Site Area Emergency An automatic or manual trip did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met AND All actions taken to shut down the reactor are not successful as indicated by reactivity control Safety Function Status Check acceptance criteria not met AND EITHER:
* REP CETs >1200&deg;F
* Applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria not met Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
None Basis:
This EAL addresses a failure of the RPS to initiate or complete an automatic reactor trip or failure of a manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor 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.
Reactor shutdown achieved by use of other trip actions such as opening supply breakers, emergency boration, or manually driving control rods are also credited as a successful manual trip provided reactivity control Safety Function Status Check acceptance criteria are met before indications of an extreme challenge to either core cooling or heat removal exist.
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 Category F ICs/EALs. This is appropriate in that the Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shut down the reactor.
Page 219 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 5).
Failure to meet the applicable HR acceptance criteria indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink).
There are three defined and potentially applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria conditions (ref. 4):
* HR SG Heat Sink Without SIAS
* HR SG Heat Sink With SIAS
* HR Once-Through-Cooling Escalation of the emergency classification level would be via IC RG1 or FG1.
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. EPIP-11 Core Damage Assessment
: 6. NEI 99-01 SS5 Page 220 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               S - System Malfunction Subcategory:            7 - Loss of Communications Initiating Condition:   Loss of all onsite or offsite communications capabilities EAL:
SU7.1          Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 State and local agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4     Communication Methods State/
System                        Onsite                NRC Local Plant Radio System                                  X Commercial Telephone System                         X         X          X Plant Address (PA) System                           X Federal Telephone System (ENS)                                           X EMnet                                                         X Hot Ring Down (HRD) Telephone System                          X Shift Manager Cell Phones                                     X          X Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Page 221 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to State and local agencies 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.).
The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the State of Florida and local communities.
The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
This hot condition EAL is equivalent to the cold condition EAL CU5.1.
Reference(s):
: 1. St. Lucie Plant Radiological Emergency Plan
: 2. NEI 99-01 SU6 Page 222 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 8 - Containment Failure Initiating Condition:        Failure to isolate containment or loss of containment pressure control EAL:
SU8.1             Unusual Event Any penetration is not closed within 15 min. of a VALID CIS[CIAS] actuation signal OR CTMT pressure > 10 psig [5.4 psig] with < one full train of CTMT heat removal systems operating per design for  15 min.
(Notes 1, 11)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 11: One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700] gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550
[2,700] gpm Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.
Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For the first condition, the containment isolation signal (CIS[CIAS]) must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
Page 223 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design (ref. 1, 2).
One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700]
gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550 [2,700] gpm The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or
) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Section 7.3 Engineered Safety Features Actuation System
: 3. NEI 99-01 SU7 Page 224 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 9 - Hazardous Event Affecting Safety Systems Initiating Condition:        Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:
SA9.1            Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9:   If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.
Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
Table S-5      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 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 Page 225 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events 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 fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
Basis:
This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.
Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/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.
VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make Page 226 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
An event affecting equipment common to two or more trains of a safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under this EAL, as appropriate to the plant mode.
By affecting the functionality of multiple trains of a safety system, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under this EAL because the two-train impact criteria that underlie the EALs and bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/EC judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under this EAL, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and bases, and is warranted because the event was severe enough to affect the functionality of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.
Escalation of the emergency classification level would be via IC FS1 or RS1.
This hot condition EAL is equivalent of the cold condition EAL CA6.1.
Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. 0-AOP-117.02 Earthquake
: 3. 1[2]-AOP-100.01 Response to Fire
: 4. EP FAQ 2016-002
: 5. NEI 99-01 SA9 Page 227 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases

===Background===
NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.
These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:
The site-specific list of plant rooms or areas with entry-related mode applicability identified should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.
The list should 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).
Further, as specified in IC HA5:
The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.
Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.
Page 228 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases PSL Table R-3 and H-2 Bases A review of St. Lucie general operating procedures identified the following mode dependent in-plant locations where procedurally defined actions are required for normal plant operation, shutdown, and cool-down:
MODE 1 (Power Operation)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Water Treatment Plant/Storage Tanks MODE 2 (Startup)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Water Treatment Plant/Storage Tanks MODE 3 (Hot Standby)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings Page 229 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Water Treatment Plant/Storage Tanks MODE 4 (Hot Shutdown)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Containment
* Water Treatment Plant/Storage Tanks Mode 5 (Cold Shutdown)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Containment Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the external release of a hazardous gas (1[2]UFSAR Section 6.4 Habitability Systems and Section 9.4.1 Control Room Ventilation). Therefore, the Control Room is not included in this assessment or in Table H-2. IC RA3 Example EAL #2 is adequately bounded by IC RA2 Example EAL #1. Therefore, the Control Room is not included in this assessment or in Table R-3.
Page 230 of 231

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases Table R-3 & H-2 Results Note 15:   The Intake Structure includes the Intake Cooling Waiter (ICW) system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Room/Area                                       Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                          1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and                              1, 2, 3, 4, 5 Control Rooms)
Cable Spreading Room                                                      1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                                    1, 2, 3, 4, 5 Steam Trestle Area                                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                                1, 2, 3, 4, 5 Component Cooling Water Area                                              1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                      1, 2, 3, 4, 5 Refueling Water Tank                                                      1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                         1, 2, 3, 4 Containment                                                                    4, 5 Procedures Reviewed
* 1[2]-GOP-101, REACTOR OPERATING GUIDELINES DURING STEADY STATE AND SCHEDULED
* 1[2]-GOP-123, TURBINE SHUTDOWN - FULL LOAD TO ZERO LOAD
* 1[2]-GOP-201, REACTOR PLANT STARTUP - MODE 2 TO MODE 1
* 1[2]-GOP-203, REACTOR SHUTDOWN
* 1[2]-GOP-302, REACTOR PLANT STARTUP - MODE 3 TO MODE 2
* 1[2]-GOP-303, REACTOR PLANT HEATUP - MODE 3 < 1750 TO MODE 3 > 1750
* 1[2]-GOP-365, REFUELING SEQUENCING GUIDELINES
* 1[2]-GOP-403, REACTOR PLANT HEATUP - MODE 4 TO MODE 3
* 1[2]-GOP-502, DATA SHEETS REQUIRED FOR HEATUP
* 1[2]-GOP-504, REACTOR PLANT HEATUP - MODE 5 TO MODE 4
* 1[2]-AOP-22.01, RAPID DOWNPOWER Page 231 of 231

St. Lucie Plant                                         L-2019-111 Emergency Action Level Technical Bases Document       Attachment 3 Emergency Action Level Technical Bases Document

St. Lucie Plant Emergency Action Level Technical Bases Document Table of Contents

==1.0   INTRODUCTION==
 ....................................................................................................... 3 2.0   DISCUSSION............................................................................................................. 3 2.1   Background ...................................................................................................... 3 2.2   Fission Product Barriers ................................................................................... 4 2.3   Fission Product Barrier Classification Criteria .................................................. 4 2.4   EAL Organization ............................................................................................. 4 2.5   Technical Bases Information ............................................................................ 7 2.6   Operational Mode Applicability ......................................................................... 8 3.0   GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................. 9 3.1   General Considerations .................................................................................... 9 3.2   Classification Methodology ............................................................................. 10

==4.0   REFERENCES==
 ........................................................................................................ 14 4.1   Developmental................................................................................................ 14 4.2   Implementing .................................................................................................. 14 5.0   DEFINITIONS, ACRONYMS & ABBREVIATIONS .................................................. 15 5.1   Definitions ....................................................................................................... 15 5.2   Abbreviations/Acronyms ................................................................................. 20 6.0   PSL-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE........................................... 23 7.0   ATTACHMENTS ...................................................................................................... 27 7.1   Attachment 1, Emergency Action Level Technical Bases............................... 27 7.2   Attachment 2, Safe Operation & Shutdown Areas Tables R-5 & H-2 Bases .. 27 Category R - Abnormal Rad Levels / Rad Effluent............................................................. 28 Category C - Cold Shutdown / Refueling System Malfunction ........................................... 73 Category E - Independent Spent Fuel Storage Installation (ISFSI).................................. 110 Category F - Fission Product Barrier Degradation ........................................................... 113 Category H - Hazards and Other Conditions Affecting Plant Safety ................................ 160 Category S - System Malfunction .................................................................................... 205 Page 2 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document

==1.0    INTRODUCTION==

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the NEI 99-1 Revision 6 EAL Upgrade Project for St. Lucie Plant (PSL). It should be used to facilitate review of the PSL EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EPIP-01, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation.
This information may assist the Emergency Coordinator (EC) in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.
The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.
Because the information in a basis document can affect emergency classification decision-making (e.g., the EC refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).
2.0    DISCUSSION 2.1    Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the Saint Lucie Plant Emergency Plan.
In 1992, the NRC endorsed NUMARC/NESP-007 Methodology for Development of Emergency Action Levels as an alternative guidance to the original Standard Review Plan and NUREG-0654 EAL schemes.
NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:
* Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
* Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs).
* Simplifying the fission product barrier EAL threshold for a Site Area Emergency.
Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).
Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ref. 4.1.1), PSL conducted an EAL implementation upgrade project that produced the EALs discussed herein.
Page 3 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 2.2     Fission Product Barriers Fission product barrier 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.
Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies a greater probability of barrier loss and reduced certainty of maintaining the barrier.
The primary fission product barriers are:
A. Fuel Clad Barrier (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System Barrier (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment Barrier (CTMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.
2.3     Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriers and loss or potential loss of the third barrier 2.4     EAL Organization The PSL EAL scheme includes the following features:
* Division of the EAL set into three broad groups:
Page 4 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document o EALs applicable under any plant operational modes - This group would be reviewed by the EAL-user any time emergency classification is considered.
o EALs applicable only under hot operational modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.
o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode.
The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.
* Within each group, assignment of EALs to categories and subcategories:
Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The PSL EAL categories are aligned to and represent the NEI 99-01 Recognition Categories. Subcategories are used in the PSL scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The PSL EAL categories and subcategories are listed below.
The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachment 1 of this document for such information.
Page 5 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document EAL Groups, Categories and Subcategories EAL Group/Category                                EAL Subcategory Any Operating Mode:
R - Abnormal Rad Levels / Rad Effluent    1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions          1 - Security Affecting Plant Safety                2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - EC Judgment E - Independent Spent Fuel Storage        1 - Confinement Boundary Installation (ISFSI)
Hot Conditions:
S - System Malfunction                    1 - Loss of Emergency AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation   None Cold Conditions:
C - Cold Shutdown / Refueling System      1 - RCS Level Malfunction                           2 - Loss of Emergency AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems Page 6 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 2.5      Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:
Category Letter & Title Subcategory Number & Title Initiating Condition (IC)
Site-specific description of the generic IC given in NEI 99-01 Rev. 6.
EAL Identifier (enclosed in rectangle)
Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:
: 1. First character (letter): Corresponds to the EAL category as described above (R, C, E, F, H or S)
: 2. Second character (letter): The emergency classification (G, S, A or U)
G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event
: 3. Third character (number): Subcategory number within the given category.
Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).
: 4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).
Classification (enclosed in rectangle):
General Emergency (G), Site Area Emergency (S), Alert (A) or Unusual Event (U).
EAL (enclosed in rectangle)
Exact wording of the EAL as it appears in the EAL Classification Matrix.
Page 7 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 -
Cold Shutdown, 6 - Refueling, DEF - Defueled, or All. (See Section 2.6 for operating mode definitions)
Definitions:
If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.
Basis:
An EAL basis section that provides PSL-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.
Reference(s):
Source documentation from which the EAL is derived 2.6      Operational Mode Applicability THERMAL MODE                   K eff       POWER
* TAVG
: 1. Power Operation               0.99         > 5%                  325&deg;F
: 2. Startup                       0.99          5%                  325&deg;F
: 3. Hot Standby                  < 0.99           0                   325&deg;F
: 4. Hot Shutdown                 < 0.99           0            > 200&deg;F & < 325&deg;F
: 5. Cold Shutdown                < 0.99           0                   200&deg;F
: 6. Refueling **                  0.95           0                   140&deg;F Defueled                       NA            NA          NA - no fuel in reactor vessel
* Excluding Decay Heat
               **  Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
Page 8 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 3.0      GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1      General Considerations When making an emergency classification, the EC must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operational Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier thresholds.
3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8).
3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the EC should not wait until the applicable time has elapsed, but 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 radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
Page 9 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10CFR 50.72 (ref.
4.1.4).
3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording 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 threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
3.1.6 Emergency Coordinator (EC) Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the EC with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The EC will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.
3.2     Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process clock started.
When assessing an EAL that specifies a time duration for the potentially classifiable condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref.
4.1.8).
Page 10 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:
* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency should be declared.
There is no additive effect from multiple EALs meeting the same ECL. For example:
* If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).
3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.
3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the EC must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the EC, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.
Page 11 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).
3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram.
3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response - In instances in which an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:
An ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.
It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event.
Emergency classification assessments must be deliberate and timely, with no undue delays.
The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the EC completing the review Page 12 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.
In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10CFR 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).
Page 13 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document

==4.0   REFERENCES==

4.1   Developmental 4.1.1    NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2    RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.
4.1.3    NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4    10 CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5    10 CFR 50.73 License Event Report System 4.1.6    Technical Specifications Table 1.2, Operational Modes 4.1.7    C-200 Offsite Dose Calculation Manual (ODCM) 4.1.8    NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9    St. Lucie Plant Radiological Emergency Plan 4.1.10 Certificate of Compliance Appendix A NUHOMS HD System Generic Technical Specifications 4.1.11 UFSAR Chapter 9.1 Fuel Storage and Handling 4.1.12 1(2)-GMM-68.02 Emergency Closure of Containment Penetrations, Personnel Hatch and Equipment Hatch 4.2   Implementing 4.2.1 EPIP-01, Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to PSL EAL Comparison Matrix 4.2.3 PSL EAL Matrix Page 14 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 5.0     DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1     Definitions (ref. 4.1.1 except as noted)
Selected terms used in Initiating Condition, Emergency Action Level statements and EAL bases are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.
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.
CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC) (ref. 4.1.10).
CONTAINMENT CLOSURE The procedurally defined conditions or actions taken to secure containment (Primary or Secondary) and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist (ref.
4.1.12):
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
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.
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:
* Notification of Unusual Event (NOUE)
* Alert
* Site Area Emergency (SAE)
* General Emergency (GE)
EXPLOSION Page 15 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 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 require a post-event inspection to determine if the attributes of an explosion are present.
FAULTED The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
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.
FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
FLOODING A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
GENERAL EMERGENCY 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 PAG exposure levels offsite for more than the immediate site area.
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 NPP PSL 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 NPPPSL. 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).
Page 16 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document HOSTILE FORCE One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
IMMINENT The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.
IMPEDE(D)
Personnel access to a room or area is hindered to an extent that 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).
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.
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.
Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.
OWNER CONTROLLED AREA (OCA)
That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL (ref. 4.1.9).
PROJECTILE An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency (ref. 4.1.9).
REFUELING PATHWAY Refueling cavity, fuel transfer canal, and spent fuel pool (SFP), but not including the reactor vessel, comprise the refueling pathway (ref. 4.1.12).
RUPTURED The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Page 17 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
SITE AREA EMERGENCY 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 PAG exposure levels beyond the SITE BOUNDARY.
SITE BOUNDARY That line beyond which the land or property is not owned, leased or otherwise controlled by PSL (ref. 4.1.7).
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.
UNUSUAL EVENT Events are in progress or have occurred which indicate a potential degradation in 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.
VALID An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct Page 18 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
VISIBLE DAMAGE Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
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.
Page 19 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 5.2        Abbreviations/Acronyms
&deg;F ....................................................................................................... Degrees Fahrenheit
&deg; ........................................................................................................................... Degrees AC ....................................................................................................... Alternating Current AFW....................................................................................................Auxiliary Feedwater AOP ................................................................................. Abnormal Operating Procedure ARM.............................................................................................. Area Radiation Monitor ATWS ...................................................................... Anticipated Transient Without Scram CDE .......................................................................................Committed Dose Equivalent CET ............................................................................................ Core Exit Thermocouple CFR ..................................................................................... Code of Federal Regulations CIAS .....................................................................Containment Isolation Actuation Signal CPM..................................................................................................... Counts Per Minute CTMT............................................................................................................. Containment DEF .....................................................................................................................Defueled DBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current D/G ......................................................................................................... Diesel Generator EAL ............................................................................................. Emergency Action Level EC ............................................................................................... Emergency Coordinator ECCS............................................................................ Emergency Core Cooling System ECL ................................................................................. Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency FAA.................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FC ........................................................................................................... Fuel Clad Barrier FEMA............................................................... Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency GPM ....................................................................................................Gallons Per Minute HR .............................................................................................................. Heat Removal Hr. .............................................................................................................................. Hour HSM.........................................................................................Horizontal Storage Module IC ......................................................................................................... Initiating Condition ICC ............................................................................................ Inadequate Core Cooling ISFSI............................................................ Independent Spent Fuel Storage Installation Page 20 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Keff ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPSI .................................................................................... Low Pressure Safety Injection LRW........................................................................................................ Liquid Radwaste LWR................................................................................................... Light Water Reactor MCB.................................................................................................... Main Control Board MG ........................................................................................................... Motor Generator Min. ......................................................................................................................... Minute MPH........................................................................................................... Miles Per Hour mR, mRem, mrem, mREM .............................................. milli-Roentgen Equivalent Man MW .................................................................................................................... Megawatt NEI .............................................................................................. Nuclear Energy Institute NEIC ................................................................... National Earthquake Information Center NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................ Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD................................................... North American Aerospace Defense Command (NO)UE ................................................................................ Notification of Unusual Event OBE ...................................................................................... Operating Basis Earthquake OCA ...............................................................................................Owner Controlled Area ODCM............................................................................ Off-site Dose Calculation Manual ORO ................................................................................. Offsite Response Organization OTC ...................................................................................... Once-Through Core Cooling PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PORV .................................................................................. Power Operated Relief Valve PSIG ................................................................................ Pounds per Square Inch Gauge PSL ............................................................................................................. St. Lucie Plant PTS........................................................................................ Pressurized Thermal Shock R ........................................................................................................................ Roentgen RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RPS ........................................................................................ Reactor Protection System RVLMS ..............................................................Reactor Vessel Level Monitoring System SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus Page 21 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document SFP.......................................................................................................... Spent Fuel Pool SFSC .................................................................................. Safety Function Status Check SG ......................................................................................................... Steam Generator SI .............................................................................................................. Safety Injection SIAS ............................................................................. Safety Injection Actuation System SM ............................................................................................................... Shift Manager SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SRV ..................................................................................................... Safety Relief Valve TC (T/C)...................................................................................................... Thermocouple TEDE ............................................................................... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TS ................................................................................................Technical Specifications TSC .......................................................................................... Technical Support Center UFSAR ...................................................................Updated Final Safety Analysis Report USGS ............................................................................ United States Geological Survey Page 22 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 6.0    PSL-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a PSL EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the PSL EALs based on the NEI guidance can be found in the EAL Comparison Matrix.
PSL             NEI 99-01 Rev. 6 Example EAL              IC EAL RU1.1            AU1           1 RU1.2            AU1           2 RU1.3            AU1           3 RU2.1            AU2           1 RA1.1            AA1           1 RA1.2            AA1           2 RA1.3            AA1           3 RA1.4            AA1           4 RA2.1            AA2           1 RA2.2            AA2           2 RA2.3            AA2           3 RA3.1            AA3           1 RA3.2            AA3           2 RS1.1            AS1           1 RS1.2            AS1           2 RS1.3            AS1           3 RS2.1            AS2           1 RG1.1            AG1           1 RG1.2            AG1           2 RG1.3            AG1           3 RG2.1            AG2           1 Page 23 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document PSL         NEI 99-01 Rev. 6 Example EAL          IC EAL CU1.1        CU1           1 CU1.2        CU1           2 CU2.1        CU2           1 CU3.1        CU3           1 CU3.2        CU3           2 CU4.1        CU4           1 CU5.1        CU5        1, 2, 3 CA1.1        CA1           1 CA1.2        CA1           2 CA2.1        CA2           1 CA3.1        CA3         1, 2 CA6.1        CA6           1 CS1.1        CS1           1 CS1.2        CS1           2 CS1.3        CS1           3 CG1.1        CG1           1 CG1.2        CG1           2 EU1.1        EU1           1 FA1.1        FA1           1 FS1.1        FS1           1 FG1.1        FG1           1 HU1.1        HU1        1, 2, 3 HU2.1        HU2           1 HU3.1        HU3           1 Page 24 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document PSL         NEI 99-01 Rev. 6 Example EAL          IC EAL HU3.2        HU3           2 HU3.3        HU3           3 HU3.4        HU3           4 HU4.1        HU4           1 HU4.2        HU4           2 HU4.3        HU4           3 HU4.4        HU4           4 HU7.1        HU7           1 HA1.1        HA1         1, 2 HA5.1        HA5           1 HA6.1        HA6           1 HA7.1        HA7           1 HS1.1        HS1           1 HS6.1        HS6           1 HS7.1        HS7           1 HG7.1        HG7           1 SU1.1        SU1           1 SU3.1        SU2           1 SU4.1        SU3           1 SU5.1        SU4        1, 2, 3 SU6.1        SU5           1 SU6.2        SU5           2 SU7.1        SU6        1, 2, 3 SU8.1        SU7         1, 2 Page 25 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document PSL         NEI 99-01 Rev. 6 Example EAL          IC EAL SA1.1        SA1          1 SA3.1        SA2          1 SA6.1        SA5          1 SA9.1        SA9          1 SS1.1        SS1          1 SS2.1        SS8          1 SS6.1        SS5          1 SG1.1        SG1          1 SG2.1        SG8          1 Page 26 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document 7.0   ATTACHMENTS 7.1   Attachment 1, Emergency Action Level Technical Bases 7.2   Attachment 2, Safe Operation & Shutdown Areas Tables R-5 & H-2 Bases Page 27 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category R - Abnormal Rad Levels / Rad Effluent EAL Group:       ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
Events of this category pertain to the following subcategories:
: 1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
: 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.
: 3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.
Page 28 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:
RU1.1                Unusual Event Reading on any Table 1[2]R-1 effluent radiation monitor > column "UE" for  60 min.
(Notes 1, 2, 3)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 29 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
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.
Page 30 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (ref. 1, 2, 3).
EAL #2 - 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).
EAL #3 - 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 AA1RA1.
Reference(s):
: 1. C-200 Offsite Dose Calculation Manual (ODCM)
: 2. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 3. NEI 99-01 AU1 Page 31 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:
RU1.2            Unusual Event Reading on any effluent radiation monitors > 2 x the alarm setpoint established by a current radioactivity discharge permit for  60 min. (Notes 1, 2, 3)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:  If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:  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.
Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Page 32 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
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 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways .
EAL #2 - 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 Such releases are typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas) (ref. 1).
EAL #3 - 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 AA1RA1.
Reference(s):
: 1. C-200 Offsite Dose Calculation Manual (ODCM)
: 2. NEI 99-01 AU1 Page 33 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    R - Abnormal Rad Levels / Rad Effluent Subcategory:                 1 - Radiological Effluent Initiating Condition:        Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
EAL:
RU1.3            Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate
> 2 x ODCM limits for  60 min. (Notes 1, 2)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:  If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
None Basis:
This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
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.
Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
Page 34 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - 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).
EAL #3 - 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 ocean water systems, etc.).
Escalation of the emergency classification level would be via IC AA1RA1.
Reference(s):
: 1. C-200 Offsite Dose Calculation Manual (ODCM)
: 2. NEI 99-01 AU1 Page 35 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.1                Alert Reading on any Table 1[2]R-1 effluent radiation monitor > column "ALERT" for  15 min.
(Notes 1, 2, 3, 4)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Note 4         The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 36 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
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 1,000 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.
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.
Page 37 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The values of the gaseous effluent thresholds in Tables 1R-1 and 2R-1 represent 10% of the SAE values calculated in EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values (ref.
1).
Escalation of the emergency classification level would be via IC AS1RS1.
Reference(s):
: 1. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 2. NEI 99-01 AA1 Page 38 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               1 - Radiological Effluent Initiating Condition:      Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.2          Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:  The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
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.
Escalation of the emergency classification level would be via IC AS1RS1.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. NEI 99-01 AA1 Page 39 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.3             Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
This EAL is assessed per the ODCM (ref. 2)
Escalation of the emergency classification level would be via IC AS1RS1.
Page 40 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. C-200 Offsite Dose Calculation Manual (ODCM)
: 3. NEI 99-01 AA1 Page 41 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:
RA1.4             Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 10 mR/hr expected to continue for  60 min.
Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.
(Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
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 Page 42 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification level would be via IC AS1RS1.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. EPIP-10 Off-site Radiological Monitoring
: 3. NEI 99-01 AA1 Page 43 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.1                Site Area Emergency Reading on any Table 1[2]R-1 effluent radiation monitor > column "SAE" for  15 min.
(Notes 1, 2, 3, 4)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Note 4:        The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 44 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
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 1,000 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.
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.
Escalation of the emergency classification level would be via IC AG1RG1.
Page 45 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 2. NEI 99-01 AS1 Page 46 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   R - Abnormal Rad Levels / Rad Effluent Subcategory:                1 - Radiological Effluent Initiating Condition:       Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.2           Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:   The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
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.
Escalation of the emergency classification level would be via IC AG1RG1.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. NEI 99-01 AS1 Page 47 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RS1.3             Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 100 mR/hr expected to continue for  60 min.
Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.
(Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 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.
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.
Escalation of the emergency classification level would be via IC AG1RG1.
Page 48 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. EPIP-10 Off-site Radiological Monitoring
: 3. NEI 99-01 AS1 Page 49 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                          R - Abnormal Rad Levels / Rad Effluent Subcategory:                       1 - Radiological Effluent Initiating Condition:              Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.1                General Emergency Reading on any Table 1[2]R-1 effluent radiation monitor > column "GE" for  15 min.
(Notes 1, 2, 3, 4)
Note 1:        The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:        If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Note 3:        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.
Note 4:        The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor               GE                SAE                Alert               UE Plant Vent 1.55E+02 &#xb5;Ci/cc    1.55E+01 &#xb5;Ci/cc    1.55E+00 &#xb5;Ci/cc    2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc    7.84E+00 &#xb5;Ci/cc    7.84E-01 &#xb5;Ci/cc          N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc    2.23E+02 &#xb5;Ci/cc    2.23E+01 &#xb5;Ci/cc    9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A                N/A                N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A                N/A                N/A                N/A           2.44E+04 cpm B" SG RE-26-45-1B                N/A                N/A                N/A           5.81E+04 cpm Page 50 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
               - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
               - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B Mode Applicability:
All Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
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 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
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.
Reference(s):
: 1. EP-CALC-PSL-1901, Rev. 0 Radiological Effluent EAL Values
: 2. NEI 99-01 AG1 Page 51 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               1 - Radiological Effluent Initiating Condition:      Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.2          General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4:  The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
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.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. NEI 99-01 AG1 Page 52 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  1 - Radiological Effluent Initiating Condition:         Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RG1.3             General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 1,000 mR/hr expected to continue for  60 min.
Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.
(Notes 1, 2)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 2:   If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
Mode Applicability:
All Definition(s):
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.
Page 53 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Reference(s):
: 1. EPIP-09 Off-site Dose Calculations
: 2. EPIP-10 Off-site Radiological Monitoring
: 3. NEI 99-01 AG1 Page 54 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               R - Abnormal Rad Levels / Rad Effluent Subcategory:            2 - Irradiated Fuel Event Initiating Condition:   UNPLANNED loss of water level above irradiated fuel EAL:
RU2.1        Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following:
* SFP or reactor cavity level instrument
* SFP or reactor cavity low level alarm
* Report of dropping level in SFP or reactor cavity AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table 1[2]R-2 radiation monitors Table 1R-2 Unit 1 Refueling Area Radiation Monitors ARM #         Channel #                 Description RE26-4-1             4        Containment CIS Monitor RE26-6-1             6        Containment CIS Monitor RE26-7-1             7        Spent Fuel Pool RE26-8-1             8        Refueling Canal RE26-36-1           36        Operating Deck Page 55 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-2 Unit 2 Refueling Area Radiation Monitors ARM #      RMCS Channel                  Description RIM-26-2          GAG-002       CCW RIM-26-4          GAG-004       Containment CIS Monitor B RIM-26-6          GAG-006       Containment CIS Monitor D RIM-26-7          GAG-007       Spent Fuel Pool RIM-26-8          GAG-008       Spent Fuel Pool RIM-26-9          GAG-009       Spent Fuel Pool RIM-26-10         GAG-010       Spent Fuel Pool RIM-26-11         GAG-011       Spent Fuel Pool RIM-26-12         GAG-012       Spent Fuel Pool RIM-26-33         GAG-033       Refueling Canal Mode Applicability:
All Definition(s):
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- Refueling cavity, fuel transfer canal, and spent fuel pool (SFP), but not including the reactor vessel, comprise the refueling pathway.
Basis:
This IC addresses a decrease 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 decrease 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). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
Page 56 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The specified area radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 1, 2). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING PATHWAY level are not classifiable under this EAL.
The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase 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 AA2RA2.
Reference(s):
: 1. 1[2]-AOP-04.01 Fuel Pool Cooling System
: 2. PSL RM EAL DBD Radiation Monitor EAL Values Used In Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 AU2 Page 57 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.1          Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC).
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.
REFUELING PATHWAY- Refueling cavity, fuel transfer canal, and spent fuel pool (SFP), but not including the reactor vessel, comprise the refueling pathway.
Basis:
This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel poolREFUELING PATHWAY (see Developer Notes). 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 EAL 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.
Escalation of the emergency would be based on either Recognition Category A R or C ICsEALs.
EAL #1 This EAL escalates from AU2 RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase 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 Page 58 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes..EAL #2 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 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 classification level would be via ICs AS1 or AS2 (see AS2 Developer Notes).
Reference(s):
: 1. NEI 99-01 AA2 Page 59 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               R - Abnormal Rad Levels / Rad Effluent Subcategory:            2 - Irradiated Fuel Event Initiating Condition:   Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.2         Alert Damage to irradiated fuel resulting in a release of radioactivity AND VALID HIGH alarm on any of the following:
* Any Table 1[2]R-2 area radiation monitors
* RSC 26-4 [RSC-26-12] Fuel Handling Building Exhaust Vent Table 1R-2 Unit 1 Refueling Area Radiation Monitors ARM #         Channel #                  Description RE26-4-1             4         Containment CIS Monitor RE26-6-1             6         Containment CIS Monitor RE26-7-1             7         Spent Fuel Pool RE26-8-1             8         Refueling Canal RE26-36-1           36         Operating Deck Page 60 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table 2R-2 Unit 2 Refueling Area Radiation Monitors ARM #     RMCS Channel                  Description RIM-26-2         GAG-002       CCW RIM-26-4         GAG-004       Containment CIS Monitor B RIM-26-6         GAG-006       Containment CIS Monitor D RIM-26-7         GAG-007       Spent Fuel Pool RIM-26-8         GAG-008       Spent Fuel Pool RIM-26-9         GAG-009       Spent Fuel Pool RIM-26-10        GAG-010       Spent Fuel Pool RIM-26-11        GAG-011       Spent Fuel Pool RIM-26-12        GAG-012       Spent Fuel Pool RIM-26-33        GAG-033       Refueling Canal Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC).
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
The specified radiation monitors are those expected to see increased area radiation levels as a result of damage to irradiated fuel (ref. 1, 2).
This IC EAL addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see Developer Notes). 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 EAL applies to irradiated fuel that is licensed for dry storage up to the point that the Page 61 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase 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.
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 #3Spent 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 A R or C ICs.
Escalation of the emergency classification level would be via ICs AS1or AS2 (see AS2 Developer Notes).
Reference(s):
: 1. 1[2]-AOP-04.01 Fuel Pool Cooling System
: 2. PSL RM EAL DBD Radiation Monitor EAL Values Used In Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 AA2 Page 62 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Significant lowering of water level above, or damage to, irradiated fuel EAL:
RA2.3          Alert Lowering of spent fuel pool level to 46 ft. 3 in. [46 ft. 5 in.] (Level 2) on LI-04-1[2]A or LI 1[2]B Mode Applicability:
All Definition(s):
None Basis:
This IC EAL addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see Developer Notes). 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.
Escalation of the emergency would be based on either Recognition Category A or C ICs. EAL #This EAL escalates from AU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.
While an area radiation monitor could detect an increase 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.
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 Page 63 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 #3Spent 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 classification level would be via ICs AS1 RS1 or ARS2 (see AS2 Developer Notes).
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (LI-04-1[2]A and LI-04-1[2]B) capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 46 ft. 3 in. [46 ft. 5 in.]) and SFP level at the top of the fuel racks (Level 3 - 36 ft. 3 in. [36 ft. 5 in.]). PSL has designated Level 3 as 1 ft. above the highest point of any fuel rack (37 ft. 3 in. [37 ft. 5 in.]) (ref. 1).
Reference(s):
: 1. EC-280519 [280520] - SPF Level Inst. Per NRC Order EA-12-051
: 2. NEI 99-01 AA2 Page 64 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 R - Abnormal Rad Levels / Rad Effluent Subcategory:              2 - Irradiated Fuel Event Initiating Condition:     Spent fuel pool level at the top of the fuel racks EAL:
RS2.1          Site Area Emergency Lowering of spent fuel pool level to 37 ft. 3 in. [37 ft. 5 in.] (Level 3) on LI-04-1[2]A or LI 1[2]B Mode Applicability:
All Definition(s):
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.
Basis:
This IC EAL 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 AG1 RG1 or AG2RG2.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (LI-04-1[2]A and LI-04-1[2]B) capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 46 ft. 3 in. [46 ft. 5 in.]) and SFP level at the top of the fuel racks (Level 3 - 36 ft. 3 in. [36 ft. 5 in.]). PSL has designated Level 3 as 1 ft. above the highest point of any fuel rack (37 ft. 3 in. [37 ft. 5 in.]) (ref. 1).
Reference(s):
: 1. EC-280519 [280520] - SPF Level Inst. Per NRC Order EA-12-051
: 2. NEI 99-01 AS2 Page 65 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     R - Abnormal Rad Levels / Rad Effluent Subcategory:                  2 - Irradiated Fuel Event Initiating Condition:         Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
RG2.1             General Emergency Spent fuel pool level cannot be restored to at least 37 ft. 3 in. [37 ft. 5 in.] (Level 3) on LI-04-1[2]A or LI-04-1[2]B for  60 min. (Note 1)
Note 1: The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
None Basis:
This IC EAL 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 EAL would likely not be met until well after another General Emergency IC EAL was met; however, it is included to provide classification diversity.
Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (LI-04-1[2]A and LI-04-1[2]B) capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2 - 46 ft. 3 in. [46 ft. 5 in.]) and SFP level at the top of the fuel racks (Level 3 - 36 ft. 3 in. [36 ft. 5 in.]). PSL has designated Level 3 as 1 ft. above the highest point of any fuel rack (37 ft. 3 in. [37 ft. 5 in.]) (ref. 1).
Reference(s):
: 1. EC-280519 [280520] - SPF Level Inst. Per NRC Order EA-12-051
: 2. NEI 99-01 AG2 Page 66 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  R - Abnormal Rad Levels / Rad Effluent Subcategory:               3 - Area Radiation Levels Initiating Condition:      Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.1            Alert Dose rate > 15 mR/hr in EITHER of the following areas:
* Control Room
* Central Alarm Station Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
Basis:
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable. 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 at the time of the elevated radiation levels. 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, 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 increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
Page 67 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 AR, C or F ICs.
Reference(s):
: 1. 1[2]-AOP-26.02 Area Radiation Monitors
: 2. NEI 99-01 AA3 Page 68 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                       R - Abnormal Rad Levels / Rad Effluent Subcategory:                    3 - Area Radiation Levels Initiating Condition:           Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
RA3.2             Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-3 room or area (Note 5)
Note 5:   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.
Note 15:  The Intake Structure includes the Intake Cooling Waiter (ICW) system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table R-3 Safe Operation & Shutdown Rooms/Areas Room/Area                                        Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                          1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and                              1, 2, 3, 4, 5 Control Rooms)
Cable Spreading Room                                                      1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                                    1, 2, 3, 4, 5 Steam Trestle Area                                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                                1, 2, 3, 4, 5 Component Cooling Water Area                                              1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                      1, 2, 3, 4, 5 Refueling Water Tank                                                      1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                         1, 2, 3, 4 Containment                                                                    4, 5 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
Page 69 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Page 70 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For EAL #2 RA3.2, 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 at the time of the elevated radiation levels. 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, 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 increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).
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 AR, C or F ICs.
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
Reference(s):
Page 71 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 1. Attachment 2 Safe Operation & Shutdown Areas Tables 1R-3 & H-2 Bases
: 2. NEI 99-01 AA3 Page 72 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature  200&#xba;F); EALs in this category are applicable only in one or more cold operating modes.
Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 -
Cold Shutdown, 6 - Refueling, DEF - Defueled).
The events of this category pertain to the following subcategories:
: 1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.
: 2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV emergency buses.
: 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.
: 4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
: 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
Page 73 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of multiple safety system trains warranting classification.
Page 74 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  1 - RCS Level Initiating Condition:         UNPLANNED loss of RCS inventory EAL:
CU1.1             Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
Basis:
RCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.
With the plant in Cold Shutdown, RCS water level is normally maintained within a pressurizer level control band (ref. 1). However, if RCS level is being controlled below the normal pressurizer level control band, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.
With the plant in Refueling mode, RCS water level is normally maintained at or above the reactor vessel flange (ref. 2).
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor (reactor vessel/RCS
[PWR] or RPV [BWR]) level concurrent with indications of coolant leakage. Either of these conditions is are considered to be a potential degradation of the level of safety of the plant.
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.
This EAL #1 recognizes that the minimum required (reactor vessel/RCS [PWR] or RPV
[BWR]) 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.
Page 75 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 (reactor vessel/RCS
[PWR] or RPV [BWR]) level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]).
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
Reference(s):
: 1. 1[2]-GOP-305 Reactor Plant Cooldown - Hot Standby to Cold Shutdown
: 2. 1[2]-GOP-365 Refueling Sequencing Guidelines
: 3. NEI 99-01 CU1 Page 76 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      UNPLANNED loss of RCS inventory EAL:
CU1.2          Unusual Event RCS water level cannot be monitored AND EITHER:
* UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter changes 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.
Basis:
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor (reactor vessel/RCS [PWR]
or RPV [BWR]) 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.
Page 77 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 (reactor vessel/RCS [PWR] or [BWR])
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.
This EAL #2 addresses a condition where all means to determine (reactor vessel/RCS [PWR]
or [BWR]) 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 (Table C-1) (ref. 1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or [BWR]).
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. NEI 99-01 CU1 Page 78 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      Significant Loss of RCS inventory EAL:
CA1.1          Alert Loss of RCS inventory as indicated by RCS water level < minimum required for continued LPSI pump operation Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
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.
For this EAL #1, a lowering of RCS water level below the minimum required for continued LPSI pump operation (site-specific level) ft indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS [PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
The minimum RCS level for continued LPSI operation is (ref. 1):
* Tygon hose 29 ft. 9 1/2 in.
* LI-1103 Level Cold Condition 0%
* LI-1117 RX REFUELING WATER LEVEL (NR) 24 1/2 in.
* LI-1117-1 RX REFUELING WATER LEVEL (WR) 24 1/2 in.
Although related, this EAL 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 Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For EAL #2, the inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels.
Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV
[BWR]).
Page 79 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1 If RCS the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
Reference(s):
: 1. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 2. NEI 99-01 CA1 Page 80 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 1 - RCS Level Initiating Condition:        Significant Loss of RCS inventory EAL:
CA1.2            Alert RCS water level cannot be monitored for  15 min. (Note 1)
AND EITHER
* UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory
* Visual observation of UNISOLABLE RCS leakage Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
Page 81 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
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.
For EAL #1, a lowering of water level below (site-specific level) indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS [PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For this EAL #2, the inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) 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 (ref. 1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]) .
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.
If the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. NEI 99-01 CA1 Page 82 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  C - Cold Shutdown / Refueling System Malfunction Subcategory:               1 - RCS Level Initiating Condition:      Loss of RCS inventory affecting core decay heat removal capability EAL:
CS1.1          Site Area Emergency 0 in. on LI-1117 or LI-1117-1 Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
Basis:
This IC addresses a significant and prolonged loss of (reactor vessel/RCS [PWR] or RCS
[BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. 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 RCS level. If RCS/reactor vesselRCS level cannot be restored, fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
The inability to monitor (reactor vessel/RCS [PWR] or RCS [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RCS [BWR]).
Page 83 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 0 in. on LI-1117 or LI-1117-1 is the lowest RCS level that can be monitored in the cold condition and corresponds to the bottom of the RCS hot leg (ref. 1).
Thisese EALs addresses 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 ICs CG1 or AG1RG1.
Reference(s):
: 1. PSL UFSAR Table 7.5-2 Safety Related Display Instrumentation
: 2. NEI 99-01 CS1 Page 84 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   C - Cold Shutdown / Refueling System Malfunction Subcategory:                1 - RCS Level Initiating Condition:       Loss of RCS inventory affecting core decay heat removal capability EAL:
CS1.2           Site Area Emergency RCS level cannot be monitored for  30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-1 sump or tank level of sufficient magnitude to indicate core uncovery
* Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
* Containment High Range Radiation Monitor reading > 4.8E+04 R/hr (Refueling Mode)
* Erratic source range monitor indications Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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.
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Page 85 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Basis:
This IC addresses a significant and prolonged loss of (reactor vessel/RCS RCS [PWR] or RPV
[BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
In this 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 (reactor vessel/RCS [PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation (ref. 1). If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]) .
If the make-up rate to the RCS unexplainably increases above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory.
In the Refueling mode, as water level in the reactor vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in on-scale indications on area radiation monitors. A reading > 4.8E+04 R/hr on Containment High Range Radiation Monitors (CHRRMs) is indicative of RCS water level at the top of the core while in the Refueling mode (ref. 2).
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
Page 86 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases These This EALs addresses 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 CG1 or AG1RG1 Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 CS1 Page 87 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  1 - RCS Level Initiating Condition:         Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:
CG1.1             General Emergency RCS level cannot be monitored for  30 min. (Note 1)
AND Core uncovery is indicated by any of the following:
* UNPLANNED increase in any Table C-1 sump or tank level of sufficient magnitude to indicate core uncovery
* Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
* Containment High Range Radiation Monitor reading > 4.8E+04 R/hr (Refueling Mode)
* Erratic source range monitor indications AND Any Containment Challenge indication, Table C-2 Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 6:   If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.
Table C-1 Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank Page 88 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table C-2 Containment Challenge Indications
* CONTAINMENT CLOSURE not established (Note 6)
* CTMT hydrogen concentration > 4%
* UNPLANNED increase in CTMT pressure Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
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.
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.
Basis:
This IC addresses the inability to restore and maintain reactor vesselRCS 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 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.
The inability to monitor (reactor vessel/RCS [PWR] or RCS [BWR]) 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 Page 89 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]) (ref. 1).
If the make-up rate to the RCS unexplainably increases above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory.
In the Refueling mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in on-scale indications on area radiation monitors. A reading > 4.8E+04 R/hr on Containment High Range Radiation Monitors (CHRRMs) is indicative of RCS water level at the top of the core while in the Refueling mode (ref. 2).
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
In EAL 2.b, tThe 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.
Three conditions are associated with a challenge to containments capability to serve as an effective barrier to fission product release:
: 1. 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.
: 2. 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 of 4%). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). 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.
Page 90 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 3. Any UNPLANNED increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential challenge of CONTAINMENT CLOSURE capability.
UNPLANNED containment pressure increase indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
Thisese EALs addresses 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.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. 1(2)-GMM-68.02 Emergency Closure of Containment Penetrations, Personnel Hatch and Equipment Hatch
: 3. NEI 99-01 CG1 Page 91 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 2 - Loss of Emergency AC Power Initiating Condition:        Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:
CU2.1            Unusual Event AC power capability, Table 1[2]C-3, to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 reduced to a single power source for  15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table 1[2]C-3 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer
* 1A[2A] Auxiliary Transformer back fed via Main Transformer (if already aligned)
* 1B[2B] Auxiliary Transformer back fed via Main Transformer (if already aligned)
Onsite
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
Page 92 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
Table 1[2]C-3 provides a list of offsite and onsite AC electrical power sources credited for this EAL. Offsite AC power sources annotated (if already aligned) require more than 15 minutes to establish and therefore are only credited if the source was already aligned at the time of AC power loss (ref. 1, 2, 3).
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.
An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed or back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.
This cold condition EAL is equivalent to the hot condition EAL SA1.1.
Reference(s):
: 1. Unit 1[2]Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. NEI 99-01 CU2 Page 93 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  2 - Loss of Emergency AC Power Initiating Condition:         Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:
CA2.1             Alert Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators, provided it can be aligned within the 15 minute classification criteria.
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.
Page 94 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Escalation of the emergency classification level would be via ICs CS1 or AS1RS1.
This cold condition EAL is equivalent to the hot condition EAL RS1.1.
Reference(s):
: 1. Unit 1[2]Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. NEI 99-01 CU2 Page 95 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 C - Cold Shutdown / Refueling System Malfunction Subcategory:              3 - RCS Temperature Initiating Condition:     UNPLANNED increase in RCS temperature EAL:
CU3.1           Unusual Event UNPLANNED increase in RCS temperature to > 200&deg;F due to loss of decay heat removal capability Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
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.
Basis:
In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5 (ref. 1). If the RCS is intact, classification should be based on the RCS pressure increase criteria of CA3.1. RCS time to boil is calculated per 1[2]-AOP-03.02  (ref. 2).
This IC EAL addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit or the inability to determine RCS temperature and level,and represents a potential degradation of the level of safety of the plant (ref. 1). If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency DirectorEC should also refer to IC EAL CA3.1.
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 #1This EALThis EAL 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 Page 96 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 at or 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 increase in reactor coolant temperature depending on the time after shutdown (ref. 3).
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 CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
Reference(s):
: 1. Unit 1[2] Technical Specifications Table 1.2
: 2. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 3. 1[2]-GOP-365 Refueling Sequencing Guidelines
: 4. NEI 99-01 CU3 Page 97 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  3 - RCS Temperature Initiating Condition:         UNPLANNED increase in RCS temperature EAL:
CU3.2             Unusual Event Loss of all RCS temperature and RCS water level indication for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6- Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
Basis:
This IC EALEAL addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, andand 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 DirectorEC should also refer to IC EAL CA3.1.
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 increase in reactor coolant temperature depending on the time after shutdown.
Page 98 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases EAL #2This EALThis EAL 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 CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
Reference(s):
: 1. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 2. NEI 99-01 CU3 Page 99 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   C - Cold Shutdown / Refueling System Malfunction Subcategory:                3 - RCS Temperature Initiating Condition:       Inability to maintain plant in cold shutdown EAL:
CA3.1            Alert UNPLANNED increase in RCS temperature to > 200&deg;F for > Table C-4 duration due to a loss of RCS cooling (Notes 1, 13)
OR UNPLANNED RCS pressure increase > 10 psig (does not apply to solid plant conditions)
Note 1:     The EC should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Note 13:    If an RCS heat removal system is in operation within the applicable Table C-4 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.
Table C-4        RCS Heat-up Duration Thresholds CONTAINMENT RCS Status                                                    Heat-up Duration CLOSURE Status Intact AND not N/A                            60 min.
reduced inventory Not intact OR                            Established                        20 min.
reduced inventory Not established                      0 min.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure containment and associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Containment Closure requirements are met when all of the following conditions exist:
* The equipment hatch is closed with a minimum of four bolts.
* All other penetrations are capable of being closed by a containment automatic isolation valve system or closed by manual valve or blind flanges.
Page 100 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
Basis:
In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5 (ref. 1). If the RCS is intact, classification should be based on the RCS pressure increase criteria of CA3.1. RCS time to boil is calculated per 1[2]-AOP-03.02  (ref. 2).
The RCS is considered to be at reduced inventory when RCS level is less than or equal to 3 ft.
below the reactor vessel flange (ref. 3).
This IC EAL 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 an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresholds table also addresses an increase 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 increase without a substantial degradation in plant safety.
Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory [PWR], 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.
The RCS pressure increase threshold EAL #2 provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.
Escalation of the emergency classification level would be via IC CS1 or AS1RS1.
Reference(s):
: 1. Unit 1[2] Technical Specifications Table 1.2
: 2. 1[2]-AOP-03.02 Shutdown Cooling Abnormal Operations
: 3. 1[2]-GOP-365 Refueling Sequencing Guidelines
: 4. NEI 99-01 CA3 Page 101 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     C - Cold Shutdown / Refueling System Malfunction Subcategory:                  4 - Loss of Vital DC Power Initiating Condition:         Loss of vital DC power for 15 minutes or longer EAL:
CU4.1             Unusual Event Indicated voltage < 112 VDC on required vital 125 VDC buses 1[2]A and 1[2]B for 15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis The vital (Class 1E) 125 V DC power system consists of two physically and electrically separated redundant buses - Bus 1A and Bus 1B. Each bus has a 60 cell battery bank with an eight-hour rating of 2400 amp-hours, with an operating range of 105 Volts to 140 Volts (ref. 1).
However, the current design minimum voltage is 112 Volts for the worst case safety load to have sufficient terminal voltage (ref. 2).
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 Page 102 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 CA1 or CA3, or an IC in Recognition Category AR.
This cold condition EAL is equivalent to the hot condition EAL SS2.1.
Reference(s):
: 1. Unit 1[2] UFSAR Section 8.3.2 DC Power System
: 2. PSL-ENG-SEES-06-018
: 3. NEI 99-01 CU4 Page 103 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               C - Cold Shutdown / Refueling System Malfunction Subcategory:            5 - Loss of Communications Initiating Condition:   Loss of all onsite or offsite communications capabilities EAL:
CU5.1          Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRC communication methods Table C-5     Communication Methods State/
System                        Onsite                NRC Local Plant Radio System                                  X Commercial Telephone System                         X         X          X Plant Address (PA) System                           X Federal Telephone System (ENS)                                           X EMnet                                                         X Hot Ring Down (HRD) Telephone System                          X Shift Manager Cell Phones                                     X          X Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, DEF - Defueled Definition(s):
None Page 104 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs State and local agencies 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.).
The first EAL condition EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition EAL #2 addresses a total loss of the communications methods used to notify all OROs State and local agencies of an emergency declaration. The OROs State and local agencies referred to here are (see Developer Notes)the State of Florida and affected local communities.
The third EAL condition EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
This cold condition EAL is equivalent to the hot condition EAL SU7.1.
Reference(s):
: 1. St. Lucie Plant Radiological Emergency Plan
: 2. NEI 99-01 CU5 Page 105 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    C - Cold Shutdown / Refueling System Malfunction Subcategory:                 6 - Hazardous Event Affecting Safety Systems Initiating Condition:        Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:
CA6.1            Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
* Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
* Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9:   If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.
Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
Table C-6      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 Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Page 106 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Definition(s):
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 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 fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
Basis:
This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.
Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/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 Page 107 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases reliability of the SAFETY SYSTEM train.
VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed 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.
EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/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 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/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.
An event affecting equipment common to two or more trains of a safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under this EAL, as appropriate to the plant mode.
By affecting the functionality of multiple trains of a safety system, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under this EAL because the two-train impact criteria that underlie the EALs and bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/EC judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under this EAL, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and bases, and is warranted because the event was severe enough to affect the functionality of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.
Escalation of the emergency classification level would be via IC CS1 or AS1RS1.
This cold condition EAL is equivalent to the hot condition EAL SA9.1.
Page 108 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. 0-AOP-117.02 Earthquake
: 3. 1[2]-AOP-100.01 Response to Fire
: 4. EP FAQ 2016-002
: 5. NEI 99-01 CA6 Page 109 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.
A Notification of Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.
The PSL ISFSI is located wholly within the PSL plant PROTECTED AREA. Therefore any security event related to the ISFSI is classified under Category H1 security event related EALs.
Page 110 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 E - ISFSI Subcategory:              1 - Confinement Boundary Initiating Condition:     Damage to a loaded cask CONFINEMENT BOUNDARY EAL:
EU1.1          Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HSM) > any of the following:
* 1,600 mrem/hr on the HSM front bird screen
* 10 mrem/hr on the outside HSM door centerline
* 10 mrem/hr on the HSM end shield wall exterior Mode Applicability:
All Definition(s):
CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the PSL ISFSI, Confinement Boundary is defined as the NUHOMS Dry Shielding Canister (DSC).
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.
Basis:
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.
The existence of damage is determined by radiological survey. The specified EAL threshold values correspond to 2 times the NUHOMS Horizontal Storage Module (HSM) technical specification external surface dose rate limits (ref. 1). The technical specification multiple of 2 times, which is also used in Recognition Category A R IC AU1RU1, 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.
Page 111 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Security-related events for ISFSIs are covered under ICs HU1 and HA1.
Reference(s):
: 1. Certificate of Compliance Appendix A NUHOMS HD System Generic Technical Specifications Section 5.4 HSM-H Dose Rate Evaluation Program
: 2. NEI 99-01 E-HU1 Page 112 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200&#xba;F); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly 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 fission product barriers are:
A. Fuel Clad Barrier (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
B. Reactor Coolant System Barrier (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment Barrier (CTMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. Loss means the barrier no longer assures containment of radioactive materials. Potential Loss means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:
Alert:
Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:
Loss or potential loss of any two barriers General Emergency:
Loss of any two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:
* The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
Page 113 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
* Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
* For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 RG1 has been exceeded.
* The fission product barrier thresholds specified within a scheme reflect plant-specific PSL design and operating characteristics.
* As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the containment, an interfacing system, or outside of the containment.
The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.
* At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency DirectorEC would have more assurance that there was no immediate need to escalate to a General Emergency.
Page 114 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 Fission Product Barrier Degradation Subcategory:              N/A Initiating Condition:     Any loss or any potential loss of either Fuel Clad or RCS EAL:
FA1.1          Alert Any loss or any potential loss of EITHER Fuel Clad or RCS barrier (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad 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 Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1 Reference(s):
: 1. NEI 99-01 FA1 Page 115 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  Fission Product Barrier Degradation Subcategory:               N/A Initiating Condition:      Loss or potential loss of any two barriers EAL:
FS1.1          Site Area Emergency Loss or potential loss of any two barriers (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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.
Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.
At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:
* One barrier loss and a second barrier loss (i.e., loss - loss)
* One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
* One barrier potential loss and a second barrier potential loss (i.e., potential loss -
potential loss)
At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the would have greater assurance that escalation to a General Emergency is less IMMINENT.
Reference(s):
: 1. NEI 99-01 FS1 Page 116 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 Fission Product Barrier Degradation Subcategory:              N/A Initiating Condition:     Loss of any two barriers and loss or potential loss of third barrier EAL:
FG1.1         General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references.
At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:
* Loss of Fuel Clad, RCS and Containment Barriers
* Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
* Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
* Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier Reference(s):
: 1. NEI 99-01 FG1 Page 117 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:
A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CTMT Radiation / RCS Activity D. CTMT Integrity or Bypass E. EC Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.
The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word None is entered in the cell.
Thresholds are assigned sequential numbers within each barrier column beginning with number one.
If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.
Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.
When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification.
Page 118 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FC)                                          Reactor Coolant System Barrier (RCS)                                                   Containment Barrier (CTMT)
Category                    Loss                         Potential Loss                                  Loss                             Potential Loss                                Loss                            Potential Loss
: 1. RVLMS indicates Sensors 4            1. An automatic or manual SIAS          1. UNISOLABLE RCS or SG tube          1. A leaking or RUPTURED SG is through 8 not covered                  actuation required by EITHER:            leakage > 50 gpm                      FAULTED outside of CTMT A                                                  (Note 12)
* UNISOLABLE RCS leakage            2. Uncontrolled RCS cooldown RCS or SG                    None                                                                                                     > 100&#xba;F/hr and RCS pressure                                                                   None Tube
* SG tube RUPTURE                      and temperature are to the left of Leakage                                                                                                                             the Maximum Subcooled (PT)
Curve (EOP Figure 1A or 1B)
: 1. REP CETs > 1200&deg;F                  2. REP CETs > 700&deg;F                                                             3. Applicable RCS and Core Heat                                              1. REP CETs > 1200&deg;F Removal (HR) Safety Function B                                             3. Applicable RCS and Core Heat Removal (HR) Safety Function None Status Check acceptance criteria not met for > 15 min.
None AND Inadequate                                            Status Check acceptance criteria                                                   AND                                                                         Functional Recovery not met for > 15 min.                                                           OTC not established                                                            (1[2]-EOP-15) for RCS and Core Heat Removal AND                                                                                                                                                          Heat Removal not effective OTC not established                                                                                                                                            within 15 min. (Note 1)
: 2. CTMT high range radiation                                                 2. Any CIAS [CIS] monitor                                                                                              2. CTMT high range radiation C           monitors Ch. # 58/59                                                          Ch. # 3/4/5/6 [RIM-26-3/4/5/6]                                                                                        monitors Ch. # 58/59
[RIM-26-40/41]                                                                > 1.5E+03 mR/hr                                                                                                       [RIM-26-40/41] > 2.7E+03 R/hr CTMT                                                            None                                                                             None                                  None
                 > 1.4E+02 R/hr Radiation /
RCS      3. Coolant activity > 300 Ci/gm Activity      dose equivalent I-131
: 2. CTMT isolation (CIS [CIAS]) is      3. CTMT pressure > 42 psig required AND EITHER:                         4. CTMT hydrogen concentration
                                                                                                                                                                                                                      > 4%
D
* CTMT integrity has been lost based on EC judgment           5. CTMT pressure > 10 psig [5.4] with CTMT                      None                                   None                                   None                                     None                                                             < one full train of containment heat Integrity or
* UNISOLABLE pathway from            removal systems operating per Bypass                                                                                                                                                                          CTMT atmosphere to the             design for  15 min. (Notes 1, 11) environment exists
: 3. Indications of RCS leakage outside of CTMT E       4. Any condition in the opinion of the EC that indicates loss of the
: 4. Any condition in the opinion of the EC that indicates potential
: 3. Any condition in the opinion of the EC that indicates loss of the RCS
: 4. Any condition in the opinion of the EC that indicates potential loss of
: 4. Any condition in the opinion of the EC that indicates loss of the
: 6. Any condition in the opinion of the EC that indicates potential loss of EC          fuel clad barrier                     loss of the fuel clad barrier           barrier                                the RCS barrier                        CTMT barrier                            the CTMT barrier Judgment Page 119 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:
None Page 120 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                      Fuel Clad Category:                     A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
: 1. RVLMS indicates Sensors 4 through 8 not covered (Note 12)
Note 12: This threshold is only applicable if all RVLMS Sensors 4 through 8 are not covered inclusive.
Definition(s):
None Basis:
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
PSL includes in its Inadequate Core Cooling instrumentation a reactor vessel level monitoring system (RVLMS) that is displayed to the operators and can measure discrete reactor vessel water levels from the top of the reactor vessel head to the top of the fuel alignment plate (Sensor #8). The bottom of this instrument's span in the reactor vessel plenum is the lowest available reactor vessel level indicator and is used in this EAL to represent approximately the top of active fuel (ref. 1).
Reference(s):
: 1. Unit 1 UFSAR Section 7.5.4 Safety Assessment System/Emergency Response Data Acquisition and Display System and Qualified Safety Parameter Display System
: 2. Unit 2 UFSAR Section 7.5.4 Instrumentation for Detection of Inadequate Core Cooling
: 3. NEI 99-01 RCS or SG Tube Leakage Potential Loss 1.A Page 121 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
: 1. REP CETs > 1200&deg;F Definition(s):
None Basis:
This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 1).
Reference(s):
: 1. EPIP-11 Core Damage Assessment
: 2. NEI 99-01 Inadequate Heat Removal Loss 2.A Page 122 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Fuel Clad Category:                  B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 2. REP CETs > 700&deg;F Definition(s):
None Basis:
This reading indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
700 &deg;F is the threshold for indicating a loss of subcooling and further temperature increases will potentially result in damage fuel cladding (ref. 1).
Reference(s):
: 1. EPIP-11 Core Damage Assessment
: 2. NEI 99-01 Inadequate Heat Removal Loss 2.A Page 123 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Fuel Clad Category:                 B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 3. Both SG wide range level < 15% for  15 min.
AND OTC not established Definition(s):
None Basis:
There are three defined and potentially applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria conditions (ref. 1):
* HR SG Heat Sink Without SIAS
* HR SG Heat Sink With SIAS
* HR Once-Through-Cooling Wide range level in both steam generators less than 15% for  15 minutes without Once-Through Core Cooling (OTC) established This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink) (ref. 1, 2). The 15-minute criteria provides the appropriate time for establishment of the heat removal safety function and aligns with the required Functional Recovery safety function status check frequency (ref. 2). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS Barrier Potential Loss threshold 2.AB.3; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
Reference(s):
: 1. 1[2]-EOP-15 Functional Recovery
: 2. 1[2]-EOP-06 Total Loss of Feedwater TLOF
: 3. NEI 99-01 Inadequate Heat Removal Fuel Clad Potential Loss 2.B Page 124 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                C. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:
: 2. CTMT high range radiation monitors Ch. # 58/59 [RIM-26-40/41] > 1.4E+02 R/hr Definition(s):
None Basis:
Containment radiation monitor readings greater than the value shown indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of ~1% clad failure into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within Technical Specifications and are therefore indicative of fuel damage The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 &#xb5;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(ref. 1).
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold C.1 2 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 ECL to a Site Area Emergency.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Reference(s):
: 1. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 2. NEI 99-01 CTMT Radiation / RCS Activity FC Loss 3.A Page 125 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Fuel Clad Category:                 C. CTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:
: 3. Coolant activity > 300 Ci/gm dose equivalent I-131 Definition(s):
None Basis:
This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Reference(s):
: 1. NEI 99-01 CTMT Radiation / RCS Activity FC Loss 3.B Page 126 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            C. CTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 127 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
None Page 128 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Fuel Clad Category:            D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Page 129 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:               Fuel Clad Category:              E. EC Judgment Degradation Threat: Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates loss of the Fuel Clad barrier Definition(s):
None Basis:
This threshold addresses any other factors that are to be used by the Emergency DirectorEC in determining whether the Fuel Clad barrier is lost.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 130 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                Fuel Clad Category:               E. EC Judgment Degradation Threat: Potential Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates potential loss of the Fuel Clad barrier Definition(s):
None Basis:
This threshold addresses any other factors that are to be used by the Emergency DirectorEC in determining whether the Fuel Clad barrier is potentially lost. The Emergency DirectorEC should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 131 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Reactor Coolant System Category:                 A. RCS or S/G Tube Leakage Degradation Threat: Loss Threshold:
: 1. An automatic or manual SIAS actuation required by EITHER:
* UNISOLABLE RCS leakage
* SG tube RUPTURE Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.AA.1 will also be met.
Reference(s):
: 1. 1[2]-EOP-03 Loss of Coolant Accident (LOCA)
: 2. 1[2]-EOP-04 Steam Generator Tube Rupture SGTR
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 132 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Reactor Coolant System Category:                 A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:
: 1. UNISOLABLE RCS or SG tube leakage > 50 gpm Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI)SIAS actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer levelRCS leakage is determined to exceed 50 gpm excluding normal reductions in RCS inventory such as letdown and RCP seal leakoff (ref. 1, 2).
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a the leaking steam generator (> 50 gpm) is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.AA.1 will also be met.
Reference(s):
: 1. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. 1[2]-AOP-08.02 Steam Generator Tube Leak
: 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 133 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Reactor Coolant System Category:                A. RCS or S/G Tube Leakage Degradation Threat: Potential Loss Threshold:
: 2. Uncontrolled RCS cooldown > 100&#xba;F/hr and RCS pressure and temperature are to the left of the Maximum Subcooled (PT) Curve (EOP Figure 1A or 1B)
Definition(s):
None Basis:
This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
The maximum allowed RCS cooldown rate per Technical Specifications is 100&#xba;F/hr with cold leg temperature > 180&#xba;F [105 &#xba;F] (ref. 1).
A RCS cooldown greater than 100&deg;F/hr below 500&deg;F is the temperature that requires implementation of Pressurized Thermal Shock (PTS) guidance (ref. 2, 3).
Reference(s):
: 1. Technical Specifications Section 3/4.4.9 Pressure/Temperature Limits
: 2. 1[2]-EOP-05 Excess Steam Demand Event
: 3. 1[2]-EOP-99 Appendices/Figures/Tables/Data Sheets, Figure 1A & 1B
: 4. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B Page 134 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
None Page 135 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Reactor Coolant System Category:                 B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 3. Both SG wide range level < 15% for  15 min.
AND OTC not established Definition(s):
None Basis:
There are three defined and potentially applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria conditions (ref. 1):
* HR SG Heat Sink Without SIAS
* HR SG Heat Sink With SIAS
* HR Once-Through-Cooling Wide range level in both steam generators less than 15% for  15 minutes without Once-Through Core Cooling (OTC) established This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink) (ref. 1, 2). The 15-minute criteria provides the appropriate time for establishment of the heat removal safety function and aligns with the required Functional Recovery safety function status check frequency (ref. 2). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B4B.3; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
Reference(s):
: 1. 1[2]-EOP-15 Functional Recovery
: 2. 1[2]-EOP-06 Total Loss of Feedwater TLOF
: 3. NEI 99-01 Inadequate Heat Removal RCS Potential Loss 2.A Page 136 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Reactor Coolant System Category:                C. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:
: 2. Any CIAS [CIS] monitor Ch. # 3/4/5/6 [RIM-26-3/4/5/6] > 1.5E+03 mR/hr Definition(s):
None Basis:
A reading > the value shown is indicative of a breach in the RCS barrier (ref. 1).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits (0.1% failed fuel). This value is lower than that specified for Fuel Clad barrier loss threshold 3.AC.2 since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.
Reference(s):
: 1. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 2. NEI 99-01 CMT Radiation / RCS Activity RCS Loss 3.A Page 137 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            C. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:
None Page 138 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
None Page 139 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Reactor Coolant System Category:            D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Page 140 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:               Reactor Coolant System Category:              E. EC Judgment Degradation Threat: Loss Threshold:
: 3. Any condition in the opinion of the EC that indicates loss of the RCS barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the Emergency DirectorEC in determining whether the RCS barrier is lost.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 141 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                Reactor Coolant System Category:               E. EC Judgment Degradation Threat: Potential Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates potential loss of the RCS barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the Emergency DirectorEC in determining whether the RCS barrier is potentially lost. The Emergency DirectorEC should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 142 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Containment Category:                  A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:
: 1. A leaking or RUPTURED SG is FAULTED outside of CTMT Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss 1.A A.1and Loss 1.AA.1, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably ([part of the FAULTED definition)]
and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).
With the exception of the initial cooldown to < 510 &#xba;F per EOP-04, This this threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump (ref. 1). These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Page 143 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A R ICs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate                       Yes                     No Less than or equal to 25 gpm           No classification       No classification Unusual Event per       Unusual Event per Greater than 25 gpm SU4SU5.1                SU4SU5.1 Requires operation of a standby Site Area Emergency per charging (makeup) pump (RCS                                    Alert per FA1.1 FS1.1 Barrier Potential Loss)
Requires an automatic or manual Site Area Emergency per ECCS (SIAS) actuation (RCS                                     Alert per FA1.1 FS1.1 Barrier Loss)
There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
Reference(s):
: 1. 1[2]-EOP-04 Steam Generator Tube Rupture SGTR
: 2. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A Page 144 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Containment Category:            A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
None Page 145 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Containment Category:            B. Inadequate Heat Removal Degradation Threat: Loss None Page 146 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                      Containment Category:                     B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
: 1. REP CETs > 1200&deg;F AND Functional Recovery (1[2]-EOP-15) for RCS and Core Heat Removal not effective within 15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Definition(s):
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.
Basis:
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 1).
This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered effective if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency DirectorEC should escalate the emergency classification level to a General Emergency as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
Reference(s):
: 1. EPIP-11 Core Damage Assessment
: 2. 1[2]-EOP-15 Functional Recovery Page 147 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A Page 148 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:             Containment Category:            C. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold:
None Page 149 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                   Containment Category:                  C. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:
: 2. CTMT high range radiation monitors Ch. # 58/59 [RIM-26-40/41] > 2.7E+03 R/hr Definition(s):
None Basis:
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the 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 (ref. 1).
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.
Reference(s):
: 1. PSL RM EAL DBD Basis for Radiation Monitor Values Used in Recognition Categories RU2, RA2 and Recognition Categories F and C
: 3. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A Page 150 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Containment Category:                D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
: 2. CTMT isolation (CIS [CIAS]) is required AND EITHER:
* CTMT integrity has been lost based on EC judgment
* UNISOLABLE pathway from CTMT atmosphere to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 1.AA.1.
These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds 4.A.1 and 4.A.2. The threshold specifies (CIS [CIAS]) as the site-specific isolation signals that initiate required containment isolations to preclude fission product release from the containment (ref. 1).
4.A.1First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure.
Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency DirectorEC will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 9-F-41. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this Page 151 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A ICs.
4.A.2Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term environment includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 9-F-41. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,
containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 9-F-41. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then the second threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold 4.A.1 to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A R ICs.
Reference(s):
: 1. UFSAR Section 6.2.4 Containment Isolation System
: 2. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A Page 152 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Containment Category:                D. CTMT Integrity or Bypass Degradation Threat: Loss Threshold:
: 3. Indications of RCS leakage outside of CTMT Definition(s):
None Basis:
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold 1.AA.1 to be met.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Containment Loss Threshold A.1.
Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.
should be sufficient to determine if RCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 9-F-41. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold CNB4 Containment Loss D.2 to be met as well.
Reference(s):
: 1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.B Page 153 of 246

St Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Figure 1: Containment Integrity or Bypass Examples Inside CTMT Auxiliary Building Open valve Penetration     Damper
* J( t Ooen valve Closed Cooling Seal Cooling Page 154 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                 Containment Category:                D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
: 3. CTMT pressure > 42 psig Definition(s):
None Basis:
If containment pressure exceeds the design pressure of 42 psig (ref. 1, 2), there exists a potential to lose the containment barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Reference(s):
: 1. PSL-ENG-SEIS-01-046 Unit 1 Emergency Operating Procedure Setpoint Basis Document
: 2. PSL-ENG-SEIS-01-047 Unit 2 Emergency Operating Procedure Setpoint Basis Document
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A Page 155 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                  Containment Category:                 D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
: 4. CTMT hydrogen concentration > 4%
Definition(s):
None Basis:
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the containment barrier.
A containment hydrogen concentration of 4% conservatively represents the lowest threshold for flammability in the presence of oxygen (ref. 1).
Reference(s):
: 1. PSL-ENG-SEIS-01-046 Unit 1 Emergency Operating Procedure Setpoint Basis Document
: 2. PSL-ENG-SEIS-01-047 Unit 2 Emergency Operating Procedure Setpoint Basis Document
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B Page 156 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                      Containment Category:                     D. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
: 5. CTMT pressure > 10 psig [5.4 psig] with < one full train of CTMT heat removal systems operating per design for  15 min. (Notes 1, 11)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 11: One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700] gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550
[2,700] gpm Definition(s):
None Basis:
This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design (ref. 1, 2). The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700]
gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550 [2,700] gpm Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Section 7.3 Engineered Safety Features Actuation System
: 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C Page 157 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:               Containment Category:              E. EC Judgment Degradation Threat: Loss Threshold:
: 4. Any condition in the opinion of the EC that indicates loss of the CTMT barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the Emergency DirectorEC in determining whether the containment barrier is lost.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Containment Loss 6.A Page 158 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Barrier:                Containment Category:               E. EC Judgment Degradation Threat: Potential Loss Threshold:
: 6. Any condition in the opinion of the EC that indicates potential loss of the CTMT barrier Definition(s):
None Basis:
This threshold addresses any other factors that may be used by the Emergency DirectorEC in determining whether the containment barrier is potentially lost. The Emergency DirectorEC should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
Reference(s):
: 1. NEI 99-01 Emergency Director Judgment Containment Potential Loss 6.A Page 159 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
: 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
: 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
: 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.
: 4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown.
: 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.
: 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
: 7. EC Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the EC the latitude to classify emergency conditions consistent with the established classification criteria based upon EC judgment.
Page 160 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     Confirmed SECURITY CONDITION or threat EAL:
HU1.1           Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by PSL Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA (OCA) - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
Page 161 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
Basis:
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 &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, and HS1 and HG1.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
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].
The first threshold EAL #1 references the Security Shift Supervision (site-specific security shift supervision)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 &sect; 2.39 information.
The second threshold EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with SFI-2416 Security and Nuclear Operations Safeguards Contingencies (site-specific procedure) (ref. 1).
The third threshold EAL #3 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 SY-AA-102-1014 Threat Assessment and Reporting (ref. 2)(site-specific procedure).
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security and Nuclear Operations Safeguards Contingencies (ref. 1).
Page 162 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC HA1.
Reference(s):
: 1. SFI-2416 Security and Nuclear Operations Safeguards Contingencies
: 2. SY-AA-102-1014 Threat Assessment and Reporting
: 3. NEI 99-01 HU1 Page 163 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:
HA1.1          Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by PSL Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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 - One or more 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.
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
Page 164 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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.
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 (OROs), allowing them to be better prepared should it be necessary to consider further actions.
This IC EAL 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 &sect; 73.71 or 10 CFR &sect; 50.72.
The first threshold EAL #1 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 second threshold EAL #2 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 SY-AA-102-1014 Threat Assessment and Reporting (site-specific procedure)(ref. 2).
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.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents Page 165 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases such as the Security and Nuclear Operations Safeguards Contingencies Security Plan for (ref.
1).
Escalation of the emergency classification level would be via IC HS1.
Reference(s):
: 1. SFI-2416 Security and Nuclear Operations Safeguards Contingencies
: 2. SY-AA-102-1014 Threat Assessment and Reporting
: 3. NEI 99-01 HA1 Page 166 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards Subcategory:              1 - Security Initiating Condition:     HOSTILE ACTION within the PROTECTED AREA EAL:
HS1.1          Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by PSL Security Shift Supervision Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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 - One or more 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.
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
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 (ref. 1, 2).
Page 167 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
This IC EAL does not apply to a HOSTILE ACTION directed at an ISFSI Protected Area located outside the 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 &sect; 73.71 or 10 CFR &sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the SFI-2416 Security and Nuclear Operations Safeguards Contingencies Security Plan for (ref. 1).
Escalation of the emergency classification level would be via IC HG1.
Reference(s):
: 1. SFI-2416 Security and Nuclear Operations Safeguards Contingencies
: 2. SY-AA-102-1014 Threat Assessment and Reporting
: 3. NEI 99-01 HS1 Page 168 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              2 - Seismic Event Initiating Condition:     Seismic event greater than OBE levels EAL:
HU2.1           Unusual Event Seismic event > OBE as indicated by S-36[S-44] CNTMT SEISMIC MONITOR HIGH LIMIT alarm Mode Applicability:
All Definition(s):
None 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). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) 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., lateral accelerations in excess of 0.08g05g). The Shift Manager or Emergency DirectorEC may seek external verification if deemed appropriate (e.g., a call to the U.S. Geological Survey (USGS), check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
For both Unit 1 and Unit 2, the OBE ground acceleration threshold is > 0.05g horizontal and vertical (ref. 3). The Control Rooms have real time OBE exceedance alarm indications (S-36[S-44] CNTMT SEISMIC MONITOR HIGH LIMIT alarm) (ref. 1, 2, 3, 4).
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Reference(s):
: 1. 0-AOP-117.02 Earthquake
: 2. 1[2] S-36[S-44] CNTMT SEISMIC MONITOR HIGH LIMIT
: 3. Unit 1 USFSAR Section 2.5.2.11 Operating Basis Earthquake
: 4. Unit 2 USFSAR Section 2.5.2.7 Operating Basis Earthquake
: 4. NEI 99-01 HU2 Page 169 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.1          Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.
EAL #2 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 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 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses (site-specific description).
Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.
Page 170 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under IC CA6 or SA9.
A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.
Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. NEI 99-01 HU3 Page 171 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.2           Unusual Event 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 Mode Applicability:
All Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.
This EAL 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 (ref. 1, 2).
EAL #3 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 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 Page 172 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses (site-specific description).
Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.
Refer to EAL CA6.1 or SA9.1 for internal flooding affecting more than one SAFETY SYSTEM train.
Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather
: 2. 1[2]-AOP-24.01 RAB Flooding
: 2. NEI 99-01 HU3 Page 173 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.3          Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.
This EAL 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, operabilit of the affected component must be required by Technical Specifications for the current operating mode.
EAL #3 This EAL addresses a hazardous materials event originating at an offsite location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.
EAL #4 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.
Page 174 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses (site-specific description).
Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.
Reference(s):
: 1. NEI 99-01 HU3 Page 175 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               3 - Natural or Technological Hazard Initiating Condition:      Hazardous event EAL:
HU3.4          Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)
Note 7:  This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Mode Applicability:
All Definition(s):
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA.
This EAL 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 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.
This EAL EAL #4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011.
EAL #5 addresses (site-specific description). Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, S or C.
Page 176 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. NEI 99-01 HU3 Page 177 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                  4 - Fire Initiating Condition:         FIRE potentially degrading the level of safety of the plant EAL:
HU4.1             Unusual Event A FIRE is not extinguished within 15 min. of any of the following fire detection indications (Note 1):
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire zone alarms or indications
* Field verification of a single fire zone alarm AND The FIRE is located within any Table H-1 area Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table H-1 PSL Fire Areas
* Reactor Containment Building/Shield Building
* Control Room
* Reactor Auxiliary Building (RAB)
* Diesel Generator Building
* Component Cooling Water (CCW)
* Intake Structure
* Intake Cooling Water (ICW)
* Steam Trestle Area
* Yard Areas:
o RWT o CST o DOST Mode Applicability:
All Definition(s):
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.
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the Page 178 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
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 The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system zone alarms or field validation of a single fire zone alarm. The zone alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.
Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).
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 alarm, indication or report. EAL #2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.
If an actual FIRE is verified by a report from the field, then EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI Page 179 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area]
EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area]
PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. 1[2]-AOP-AOP-100.11 Response to Fire - Fire Zones
: 3. NEI 99-01 HU4 Page 180 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                 4 - Fire Initiating Condition:        FIRE potentially degrading the level of safety of the plant EAL:
HU4.2            Unusual Event Receipt of a single fire zone alarm (i.e., no other indications of a FIRE)
AND The fire zone alarm is indicating a FIRE within any Table H-1 area (excluding containment)
(Note 14)
AND The existence of a FIRE is not verified (proved or disproved) within 30 min. of alarm receipt (Note 1)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 14: A containment fire zone alarm is considered VALID upon receipt of multiple zones (more than one) actuated on CP-376 panel.
Table H-1 PSL Fire Areas
* Reactor Containment Building/Shield Building
* Control Room
* Reactor Auxiliary Building (RAB)
* Diesel Generator Building
* Component Cooling Water (CCW)
* Intake Structure
* Intake Cooling Water (ICW)
* Steam Trestle Area
* Yard Areas:
o RWT o CST o DOST Mode Applicability:
All Definition(s):
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.
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct Page 181 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
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 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 alarm, indication or report.
EAL #2 The 30-minute requirement begins upon receipt of a single VALID fire detection system zone alarm (excluding containment). The zone alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30-minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15-minute requirement beginning with the verification of the fire by field report.
Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1, 2).
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.
With regard to containment fire zone alarms, there is constant air movement in containment due to the operation of the CAH system. The operating cooling units are drawing air to the units past the smoke detectors. It can be reasonably expected that a fire that burns for 15 minutes would produce sufficient products of combustion to cause fire zone detectors in multiple zones to alarm. Therefore a single containment fire zone alarm is not considered VALID.
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.
Page 182 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases If an actual FIRE is verified by a report from the field, then HU4.1 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 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area]
EAL #4 Basis-Related Requirements from Appendix R (Note; While PSL is an NFP-805 plant, the following excerpt from 10CFR 50 Appendix R provides the bases for the acceptability of the use of a 30-minute validation time for a single fire zone alarm.)
Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2EAL #2, the 30-minutes to verify (prove or disprove) 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 CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. 1[2]-AOP-AOP-100.11 Response to Fire - Fire Zones
: 3. NEI 99-01 HU4 Page 183 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                  4 - Fire Initiating Condition:         FIRE potentially degrading the level of safety of the plant EAL:
HU4.3             Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
All Definition(s):
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 - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
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 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 alarm, indication or report.
EAL #2 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.
Page 184 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 In addition to a FIRE addressed by EAL HU4.1 #1 or HU4.2EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety.
This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area]EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area]
PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one Page 185 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. NEI 99-01 HU4 Page 186 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               4 - Fire Initiating Condition:      FIRE potentially degrading the level of safety of the plant EAL:
HU4.4          Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:
All Definition(s):
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 - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
Basis:
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 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 alarm, indication or report.
EAL #2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to Page 187 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an ISFSI outside the plant Protected Area]
EAL #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area]
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.
The Shift Fire Brigade Advisor or Shift Fire Brigade Leader will assess whether the fire conditions warrant outside assistance (ref. 1).
Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.
Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
Page 188 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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 CA6 or SA9.
Reference(s):
: 1. 1[2]-AOP-AOP-100.01 Response to Fire
: 2. NEI 99-01 HU4 Page 189 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                      H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                   5 - Hazardous Gases Initiating Condition:          Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:
HA5.1            Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5)
Note 5:  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.
Note 15: The Intake Structure includes the Intake Cooling Waiter (ICW) system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area                                        Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                          1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and                              1, 2, 3, 4, 5 Control Rooms)
Cable Spreading Room                                                      1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                                    1, 2, 3, 4, 5 Steam Trestle Area                                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                                1, 2, 3, 4, 5 Component Cooling Water Area                                              1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                      1, 2, 3, 4, 5 Refueling Water Tank                                                      1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                         1, 2, 3, 4 Containment                                                                    4, 5 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown Page 190 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).
Basis:
This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.
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 at the time of the gaseous release. 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 DirectorECs 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.
If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
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 Page 191 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases lead to breathing difficulties, unconsciousness or even death. Steam can be an asphyxiant if present in sufficient concentration to reduce oxygen levels below normal.
This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area , or to intentional inerting of containment. (BWR only).
Escalation of the emergency classification level would be via Recognition Category AR, C or F ICs.
Reference(s):
: 1. Attachment 2 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases
: 2. NEI 99-01 HA5 Page 192 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                H - Hazards and Other Conditions Affecting Plant Safety Subcategory:             6 - Control Room Evacuation Initiating Condition:    Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA6.1          Alert An event has resulted in the Control Room being evacuated Mode Applicability:
All Definition(s):
None 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.
The Control Room is deemed to have been evacuated when the last licensed operator leaves the Control Room.
Escalation of the emergency classification level would be via IC HS6.
Reference(s):
: 1. 1[2]-ONP-100.02 Control Room Inaccessibility
: 2. NEI 99-01 HA6 Page 193 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     H - Hazards and Other Conditions Affecting Plant Safety Subcategory:                  6 - Control Room Evacuation Initiating Condition:         Inability to control a key safety function from outside the Control Room EAL:
HS6.1             Site Area Emergency An event has resulted in the Control Room being evacuated AND Control of any of the following key safety functions is not re-established within 15 min. of the last licensed operator leaving the Control Room (Note 1):
Reactivity (modes 1, 2 and 3 only)
Core cooling RCS heat removal Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.
The determination of whether or not control is established at the remote safe shutdown location(s) is based on Emergency DirectorEC judgment. The Emergency DirectorEC is expected to make a reasonable, informed judgment within 15 (the site-specific time for transfer) minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
The Control Room is deemed to have been evacuated when the last licensed operator leaves the Control Room.
Establishment of the reactivity safety function is only applicable in Modes 1, 2 and 3. Sufficient shutdown margin has already been established once in modes 4, 5 and 6 (ref. 2).
Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):
Page 194 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 1. 1[2]-ONP-100.02 Control Room Inaccessibility
: 2. NRC EP FAQ 2015-014
: 3. NEI 99-01 HS6 Page 195 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               7 - EC Judgment Initiating Condition:      Other conditions existing that in the judgment of the EC warrant declaration of a Unusual Event EAL:
HU7.1           Unusual Event Other conditions exist which, in the judgment of the EC, 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.
Mode Applicability:
All Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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 DirectorEC to fall under the emergency classification level description for an Unusual EventNOUE.
Reference(s):
: 1. NEI 99-01 HU7 Page 196 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              7 - EC Judgment Initiating Condition:     Other conditions exist that in the judgment of the EC warrant declaration of an Alert EAL:
HA7.1          Alert Other conditions exist which, in the judgment of the EC, 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.
Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
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 DirectorEC to fall under the emergency classification level description for an Alert.
Page 197 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Page 198 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. NEI 99-01 HA7 Page 199 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  H - Hazards and Other Conditions Affecting Plant Safety Subcategory:               7 - EC Judgment Initiating Condition:      Other conditions existing that in the judgment of the EC warrant declaration of a Site Area Emergency EAL:
HS7.1           Site Area Emergency Other conditions exist which, in the judgment of the EC, 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 Mode Applicability:
All Definition(s):
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
SITE BOUNDARY - That line beyond which the land or property is not owned, leased or otherwise controlled by PSL.
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 DirectorEC to fall under the emergency classification level description for a SITE AREA EMERGENCY.
Page 200 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Page 201 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. NEI 99-01 HS7 Page 202 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 H - Hazards and Other Conditions Affecting Plant Safety Subcategory:              7 - EC Judgment Initiating Condition:     Other conditions exist that in the judgment of the EC warrant declaration of a General Emergency EAL:
HG7.1                 General Emergency Other conditions exist which, in the judgment of the EC, 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.
Mode Applicability:
All Definition(s):
HOSTAGE - A person(s) held as leverage against the station to ensure that demands will be met by the station.
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 PSL 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 PSL. 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).
OWNER CONTROLLED AREA - That portion of FPL property surrounding and including the St. Lucie Nuclear Power Plant which is subject to limited access and control as deemed appropriate by FPL.
PROJECTILE - An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA - The area (within the OWNER CONTROLLED AREA) occupied by the nuclear units and associated equipment and facilities enclosed within the security perimeter fence. The area within which accountability of personnel is maintained in an emergency.
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 Page 203 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases DirectorEC to fall under the emergency classification level description for a GENERAL EMERGENCY.
Reference(s):
: 1. NEI 99-01 HG7 Page 204 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200&#xba;F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.
The events of this category pertain to the following subcategories:
: 1. Loss of Emergency AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV emergency buses.
: 2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses.
: 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.
: 4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (~1% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
: 5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.
Page 205 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to properly result in reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.
: 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
: 8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.
: 9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system train performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.
Page 206 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 1 - Loss of Emergency AC Power Initiating Condition:        Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:
SU1.1            Unusual Event Loss of all offsite AC power capability, Table 1[2]S-1, to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. SU3.1 Table 1[2]S-1 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer Onsite
* 1A[2A] Auxiliary Transformer
* 1B[2B] Auxiliary Transformer
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Table 1[2]S-1 provides a list of offsite and onsite AC electrical power sources credited for this EAL. The onsite AC power source annotated (if already aligned) requires more than 15 minutes to establish and therefore is only credited if the source was already aligned at the time of AC power loss (ref. 1, 2, 3).
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.
Page 207 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases For emergency classification purposes, capability means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level would be via IC SA1.
Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. NEI 99-01 SU1 Page 208 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  1 - Loss of Emergency AC Power Initiating Condition:         Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:
SA1.1             Alert AC power capability, Table 1[2]S-1, to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 reduced to a single power source for  15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table 1[2]S-1 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer Onsite
* 1A[2A] Auxiliary Transformer
* 1B[2B] Auxiliary Transformer
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 209 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
Table 1[2]S-1 provides a list of offsite and onsite AC electrical power sources credited for this EAL. The onsite AC power source annotated (if already aligned) requires more than 15 minutes to establish and therefore is only credited if the source was already aligned at the time of AC power loss (ref. 1, 2, 3).
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 SU1.
An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fedcross tied to the other unit from the unit main generator.
A loss of emergency power sources (e.g., onsite diesel generators and unit cross ties) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level would be via IC SS1.
This hot condition EAL is equivalent to the cold condition EAL CU2.1.
Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. NEI 99-01 SA1 Page 210 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                   S - System Malfunction Subcategory:                1 - Loss of Emergency AC Power Initiating Condition:       Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:
SS1.1           Site Area Emergency Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
Note 1: The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators, provided it can be aligned within the 15 minute classification criteria (ref. 4, 5).
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 AG1RG1, FG1 or SG1.
This hot condition EAL is equivalent to the cold condition EAL CA2.1.
Page 211 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. 1[2]-FSG-05 FLEX Implementation
: 6. NEI 99-01 SS1 Page 212 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                  S -System Malfunction Subcategory:               1 - Loss of Vital AC Power Initiating Condition:      Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:
SG1.1           General Emergency Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 AND REP CETs > 1200&#xba;F Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators (ref. 4, 5).
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 6).
This IC addresses a prolonged loss of all power sources to AC emergency buses that results in degraded core cooling. 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 eventually lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
Page 213 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The For extended loss of emergency bus AC power events that do not result in a breach of the RCS barrier, this EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.
Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.
The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
Reference(s):
: 1. Unit 1[2]Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. 1[2]-FSG-05 FLEX Implementation
: 6. EPIP-11 Core Damage Assessment
: 7. NEI 99-01 SG1 Page 214 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  2 - Loss of Vital DC Power Initiating Condition:         Loss of all vital DC power for 15 minutes or longer EAL:
SS2.1             Site Area Emergency Indicated voltage < 112 VDC on both vital 125 VDC buses 1[2]A and 1[2]B for  15 min.
(Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
The vital (Class 1E) 125 V DC power system consists of two physically and electrically separated redundant buses - Bus 1A[2A] and Bus 1B[2B]. Each bus has a 60 cell battery bank with an eight-hour rating of 2400 amp-hours, with an operating range of 105 Volts to 140 Volts (ref. 1). However, the current design minimum voltage is 112 Volts for the worst case safety load to have sufficient terminal voltage (ref. 2).
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 AG1RG1, FG1 or SG8SG2.
This hot condition EAL equivalent of the cold condition EAL CU4.1.
Page 215 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Reference(s):
: 1. Unit 1[2] UFSAR Section 8.3.2 DC Power System
: 2. PSL-ENG-SEES-06-018
: 3. NEI 99-01 SS8 Page 216 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S -System Malfunction Subcategory:                  2 - Loss of Vital DC Power Initiating Condition:         Loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL:
SG2.1             General Emergency Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1[2]A3 and 1[2]B3 for  15 min. (Note 1)
AND Indicated voltage is < 112 VDC on both vital 125 VDC buses 1[2]A and 1[2]B for 15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
Basis:
This IC addresses a concurrent and prolonged loss of both emergency AC and vital DC power.
A loss of all emergency 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 emergency AC and vital DC power will lead to multiple challenges to fission product barriers.
For this EAL credit can be taken for any AC power source that has sufficient capability to operate equipment necessary to maintain a safe shutdown condition, such as FLEX generators (ref. 4, 5).
Page 217 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The Vital (Class 1E) 125 V DC power system consists of two physically and electrically separated redundant buses - Bus 1[2]A and Bus 1[2]B. Each bus has a 60 cell battery bank with an eight-hour rating of 2300 amp-hours, with an operating range of 105 Volts to 140 Volts (ref. 6). However, the current design minimum voltage is 112 Volts for the worst case safety load to have sufficient terminal voltage (ref. 7).
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 both EAL thresholds are met.
Reference(s):
: 1. Unit 1[2] Technical Specifications Section 3.8.1.2 Electrical Power Systems Shutdown
: 2. Unit 1[2] UFSAR section 8.2 Offsite Power System
: 3. Unit 1[2] UFSAR section 8.3 Onsite Power System
: 4. 1[2]-EOP-10 Station Blackout SBO
: 5. 1[2]-FSG-05 FLEX Implementation
: 6. Unit 1[2] UFSAR Section 8.3.2 DC Power System
: 7. PSL-ENG-SEES-06-018
: 8. NEI 99-01 SG8 Page 218 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  3 - Loss of Control Room Indications Initiating Condition:         UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1             Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for  15 min. (Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table S-2    Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CET temperature
* Level in at least one SG
* Auxiliary feedwater flow to at least one SG Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
Page 219 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
Applicable safety system parameters are listed in Table S-2.
The Safety Assessment System (SAS) and Emergency Response Data Acquisition and Display System (ERDADS) serve as redundant indicators which may be utilized as compensatory measures in lieu of the Control Room indicators associated with safety functions (ref. 1, 2).
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 one or more 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.
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 [PWR] / RPV level [BWR] and RCS heat removal.
The loss of the ability to determine one or more 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 one or more 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 [PWR] / RPVRCS water level [BWR] 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 SA2SA3.
Reference(s):
: 1. Unit 1[2] UFSAR Section 7.5 Safety Related Display Instrumentation
: 2. Unit 2 UFSAR Appendix 7.5A Safety Assessment System/Emergency Response Data Acquisition and Display System
: 3. NEI 99-01 SU2 Page 220 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 3 - Loss of Control Room Indications Initiating Condition:        UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1            Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for  15 min. (Note 1)
AND Any significant transient is in progress, Table S-3 Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Table S-2     Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CET temperature
* Level in at least one SG
* Auxiliary feedwater flow to at least one SG Table S-3    Significant Transients
* Electrical load rejection > 25% full electrical load
* Reactor Trip
* SIAS actuation Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Page 221 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Definition(s):
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
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.
Basis:
Applicable safety system parameters are listed in Table S-2.
Significant transients are listed in Table S-3.
The Safety Assessment System (SAS) and Emergency Response Data Acquisition and Display System (ERDADS) serve as redundant indicators which may be utilized as compensatory measures in lieu of the Control Room indicators associated with safety functions (ref. 1, 2).
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 one or more 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.
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 [PWR] / RPV level [BWR] and RCS heat removal.
Page 222 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases The loss of the ability to determine one or more 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 one or more 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 [PWR] / RPVRCS water level [BWR] 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 AS1RS1 Reference(s):
: 1. Unit 1[2] UFSAR Section 7.5 Safety Related Display Instrumentation
: 2. Unit 2 UFSAR Appendix 7.5A Safety Assessment System/Emergency Response Data Acquisition and Display System
: 3. NEI 99-01 SA2 Page 223 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               S - System Malfunction Subcategory:            4 - RCS Activity Initiating Condition:   Reactor coolant activity greater than Technical Specification allowable limits EAL:
SU4.1         Unusual Event Sample analysis indicates that a reactor coolant activity value is > any of the following allowable limits:
* 60 Ci/gm dose equivalent I-131 instantaneous
* 1.0 Ci/gm dose equivalent I-131 for > 48 hours
* 518.9 Ci/gm dose equivalent Xe-133 for > 48 hours Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications (ref. 1, 2, 3). This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.
Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A R ICs.
Reference(s):
: 1. Technical Specification 3.4.8 RCS Specific Activity
: 2. 1[2]-AOP-01.06 Excessive RCS Activity
: 3. CY-SL-108-005 High Activity in Reactor Coolant System
: 4. NEI 99-01 SU3 Page 224 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  5 - RCS Leakage Initiating Condition:         RCS leakage for 15 minutes or longer EAL:
SU5.1             Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for  15 min.
OR RCS identified leakage > 25 gpm for  15 min.
OR Leakage from the RCS to a location outside containment > 25 gpm for  15 min.
(Note 1)
Note 1:   The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
Once the RCS leak rate has been quantified to be greater than the specified value, failure to isolate the leak within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the time of leak rate quantification, 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.
The first and second EAL conditions EAL #1 and EAL #2 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). The third condition EAL
#3 addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions EALs thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each condition EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition 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 a relief valve does not warrant an emergency classification. For PWRs, aAn emergency classification would Page 225 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). For BWRs, 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 A R or F.
Reference(s):
: 1. Technical Specification Definitions Section 1.0
: 2. Technical Specification 3.4.6 RCS Leakage
: 3. 1[2]-AOP-01.08 RCS Leakage Abnormal Operations
: 2. NEI 99-01 SU4 Page 226 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                      S - System Malfunction Subcategory:                   6 - RPS Failure Initiating Condition:          Automatic or manual trip fails to shut down the reactor EAL:
SU6.1             Unusual Event An automatic trip did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip (RX Trip Pushbuttons) are successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria met (Note 8)
Note 8:    A manual trip 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.
Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
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.
Basis:
This IC EAL addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BWR])that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] /
scram [BWR]) 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.
Following the failure on an automatic reactor (trip[PWR] / scram [BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,
initiate a manual reactor (trip[PWR] / scram [BWR]) by pushing the RX Trip Pushbuttons). If these this manual actions are 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.
If an initial manual reactor (trip [PWR] / scram [BWR])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 (trip[PWR] / scram [BWR])) using a different switch).
Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram
[BWR]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PWR] / scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] /
scram [BWR]) 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.
Page 227 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases A manual action at the reactor control consoles console (RTGB) is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip by pushing the RX Trip Pushbuttons [PWR] / scram [BWR])). This action does not include opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally, 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.
[BWR]
The plant response to the failure of an automatic or manual reactor (trip[PWR] / scram [BWR])
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 SA5. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5 or FA1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Should a reactor (trip [PWR] / scram [BWR])signal be generated as a result of plant work (e.g.,
RPS setpoint testing), the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor (trip
[PWR] / scram [BWR])and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the (trip [PWR] / scram [BWR])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.
In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to shutdown the reactor, the event escalates to the Alert under EAL SA6.1.
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. NEI 99-01 SU5 Page 228 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                      S - System Malfunction Subcategory:                   6 - RPS Failure Initiating Condition:          Automatic or manual trip fails to shut down the reactor EAL:
SU6.2             Unusual Event A manual trip (RX Trip Pushbuttons) did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met (Note 8)
AND A subsequent automatic trip or manual trip action (RX Trip Pushbuttons) is successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria met Note 8:    A manual trip 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.
Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
None Basis:
This IC EAL addresses a failure of the RPS to initiate or complete an automatic ora manual reactor (trip [PWR] / scram [BWR])that results in a reactorto shutdown the reactor, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip[PWR] / scram [BWR]) 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.
Following the failure on an automatic reactor (trip[PWR] / scram [BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,
initiate a manual reactor (trip[PWR] / scram [BWR])). 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.
If an initial manual reactor (trip[PWR] / scram [BWR]) is unsuccessful, operators will promptly take manual actions at another other location(s) on away from the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip[PWR] / scram [BWR])) using a different switchby opening CEDM MG set feeder breakers or opening the reactor trip feeder breakers locally) (ref. 3). However, those actions are not credited as a successful manual reactor trip for this EAL.
Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] / scram
[BWR])the reactor, or a concurrent plant condition, may lead to the generation of an automatic Page 229 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases reactor (trip [PWR] / scram [BWR])signal. If a subsequent manual or automatic (trip [PWR] /
scram [BWR])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 (RTGB) is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip by pushing the RX Trip Pushbuttons [PWR] / scram [BWR])). This action does not include opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally, 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.
[BWR]
The plant response to the failure of an automatic or manual reactor (trip [PWR] / scram [BWR])
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 SA5. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5 or FA1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Should a reactor (trip [PWR] / scram [BWR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
If the signal causes a plant transient that should have included an automatic reactor (trip
[PWR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the (trip [PWR] / scram [BWR]) 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.
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. NEI 99-01 SU5 Page 230 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                     S - System Malfunction Subcategory:                  2 - RPS Failure Initiating Condition:         Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:
SA6.1            Alert An automatic or manual trip (RX Trip Pushbuttons) did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met AND Subsequent automatic or manual trip actions (RX Trip Pushbuttons) are not successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met (Note 8)
Note 8:   A manual trip 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.
Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
None Basis:
This IC EAL addresses a failure of the RPS to initiate or complete an automatic reactor trip or failure of a manual reactor (trip [PWR] / scram [BWR]) that results in a reactorto shutdown the reactor, 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 console (RTGB) is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (trip[PWR] / scram [BWR]) by pushing the RX Trip Pushbuttons). This action does not include opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locally, 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 (opening CEDM MG set feeder breakers, opening the reactor trip feeder breakers locallye.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.
Page 231 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.
[BWR]
The plant response to the failure of an automatic or manual reactor (trip [PWR] / scram [BWR])
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 shut down the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / RPV water level [BWR]or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS65. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS65 or FS1, an Alert declaration is appropriate for this event.
It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. NEI 99-01 SA5 Page 232 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                 S - System Malfunction Subcategory:              2 - RPS Failure Initiating Condition:     Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal EAL:
SS6.1          Site Area Emergency An automatic or manual trip did not shut down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met AND All actions taken to shut down the reactor are not successful as indicated by reactivity control Safety Function Status Check acceptance criteria not met AND EITHER:
* REP CETs >1200&deg;F
* Applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria not met Mode Applicability:
1 - Power Operation, 2 - Startup Definition(s):
None Basis:
This IC EAL addresses a failure of the RPS to initiate or complete an automatic reactor trip or failure of a manual reactor (trip [PWR] / scram [BWR]) that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor 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.
Reactor shutdown achieved by use of other trip actions such as opening supply breakers, emergency boration, or manually driving control rods are also credited as a successful manual trip provided reactivity control Safety Function Status Check acceptance criteria are met before indications of an extreme challenge to either core cooling or heat removal exist.
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 shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shut down the reactor.
Page 233 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria (ref. 3, 4).
Core Damage Assessment uses a CET temperature of 1200 &deg;F as the temperature indicative of a significantly damaged core (ref. 5).
Failure to meet the applicable HR acceptance criteria indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink).
There are three defined and potentially applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria conditions (ref. 4):
* HR SG Heat Sink Without SIAS
* HR SG Heat Sink With SIAS
* HR Once-Through-Cooling Escalation of the emergency classification level would be via IC AG1 RG1 or FG1.
Reference(s):
: 1. Technical Specification Table 1.2 Operational Modes
: 2. Technical Specification Table 3.3-1 Reactor Protective Instrumentation
: 3. 1[2]-EOP-01 Standard Post Trip Actions SPTA
: 4. 1[2]-EOP-15 Functional Recovery
: 5. EPIP-11 Core Damage Assessment
: 6. NEI 99-01 SS5 Page 234 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:               S - System Malfunction Subcategory:            7 - Loss of Communications Initiating Condition:   Loss of all onsite or offsite communications capabilities EAL:
SU7.1          Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 State and local agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4     Communication Methods State/
System                        Onsite                NRC Local Plant Radio System                                  X Commercial Telephone System                         X         X          X Plant Address (PA) System                           X Federal Telephone System (ENS)                                           X EMnet                                                         X Hot Ring Down (HRD) Telephone System                          X Shift Manager Cell Phones                                     X          X Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Page 235 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Basis:
This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs State and local agencies 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.).
The first EAL condition EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
The second EAL condition EAL #2 addresses a total loss of the communications methods used to notify all OROs State and local agencies of an emergency declaration. The OROs State and local agencies referred to here are the State of Florida and local communities. (see Developer Notes)
The third EAL EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
This hot condition EAL is equivalent to the cold condition EAL CU5.1.
Reference(s):
: 1. St. Lucie Plant Radiological Emergency Plan
: 2. NEI 99-01 SU6 Page 236 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 8 - Containment Failure Initiating Condition:        Failure to isolate containment or loss of containment pressure control EAL:
SU8.1             Unusual Event Any penetration is not closed within 15 min. of a VALID CIS[CIAS] actuation signal OR CTMT pressure > 10 psig [5.4 psig] with < one full train of CTMT heat removal systems operating per design for  15 min.
(Notes 1, 11)
Note 1:  The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.
Note 11: One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700] gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550
[2,700] gpm Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
This IC EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.
Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For EAL #1the first condition, the containment isolation signal (CIS[CIAS]) must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
Page 237 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases EAL #2The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design (ref. 1, 2).
One full train of containment heat removal systems consist of any one of the following:
* Two (2) Containment Spray Headers with flow in each header at least 2,550 [2,700]
gpm
* Four (4) Containment Coolers
* Two (2) Containment Coolers with at least one (1) Containment Spray Header with at least 2,550 [2,700] gpm The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
Reference(s):
: 1. UFSAR Section 6.2.2 Containment Heat Removal Systems
: 2. UFSAR Section 7.3 Engineered Safety Features Actuation System
: 3. NEI 99-01 SU7 Page 238 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases Category:                    S - System Malfunction Subcategory:                 9 - Hazardous Event Affecting Safety Systems Initiating Condition:        Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:
SA9.1            Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:
Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)
Note 9:   If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.
Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
Table S-5      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 Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
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 Page 239 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events 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 fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.
FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.
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 (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.
VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.
Basis:
This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.
Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/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.
VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make Page 240 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed 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.
EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/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 1.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/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.
An event affecting equipment common to two or more trains of a safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under this EAL, as appropriate to the plant mode.
By affecting the functionality of multiple trains of a safety system, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and bases.
An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under this EAL because the two-train impact criteria that underlie the EALs and bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/EC judgement.
An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under this EAL, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and bases, and is warranted because the event was severe enough to affect the functionality of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.
Escalation of the emergency classification level would be via IC FS1 or AS1RS1.
This hot condition EAL is equivalent of the cold condition EAL CA6.1.
Reference(s):
: 1. 0-AOP-100.14 Response to Severe Weather Page 241 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 1 Emergency Action Level Technical Bases
: 2. 0-AOP-117.02 Earthquake
: 3. 1[2]-AOP-100.01 Response to Fire
: 4. EP FAQ 2016-002
: 5. NEI 99-01 SA9 Page 242 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases

===Background===
NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.
These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:
The site-specific list of plant rooms or areas with entry-related mode applicability identified should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.
The list should 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).
Further, as specified in IC HA5:
The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.
Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.
Page 243 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases PSL Table R-3 and H-2 Bases A review of St. Lucie general operating procedures identified the following mode dependent in-plant locations where procedurally defined actions are required for normal plant operation, shutdown, and cool-down:
MODE 1 (Power Operation)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Water Treatment Plant/Storage Tanks MODE 2 (Startup)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Water Treatment Plant/Storage Tanks MODE 3 (Hot Standby)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings Page 244 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Water Treatment Plant/Storage Tanks MODE 4 (Hot Shutdown)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Containment
* Water Treatment Plant/Storage Tanks Mode 5 (Cold Shutdown)
* Reactor Auxiliary Building (0.5 and 19.5 levels)
* Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and Control Rooms)
* Cable Spreading Room
* A3/B3 Switchgear Rooms
* Steam Trestle Area
* Intake Structure
* Component Cooling Water Area
* Emergency Diesel Generator Buildings
* Diesel Oil Storage Tanks
* Refueling Water Tank
* Containment Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the external release of a hazardous gas (1[2]UFSAR Section 6.4 Habitability Systems and Section 9.4.1 Control Room Ventilation). Therefore, the Control Room is not included in this assessment or in Table H-2. IC RA3 Example EAL #2 is adequately bounded by IC RA2 Example EAL #1. Therefore, the Control Room is not included in this assessment or in Table R-3.
Page 245 of 246

St. Lucie Plant Emergency Action Level Technical Bases Document Attachment 2 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases Table R-3 & H-2 Results Note 15:   The Intake Structure includes the Intake Cooling Waiter (ICW) system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Room/Area                                       Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                          1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, Steam Trestle, Switchgear and                              1, 2, 3, 4, 5 Control Rooms)
Cable Spreading Room                                                      1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                                    1, 2, 3, 4, 5 Steam Trestle Area                                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                                1, 2, 3, 4, 5 Component Cooling Water Area                                              1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                      1, 2, 3, 4, 5 Refueling Water Tank                                                      1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                         1, 2, 3, 4 Containment                                                                    4, 5 Procedures Reviewed
* 1[2]-GOP-101, REACTOR OPERATING GUIDELINES DURING STEADY STATE AND SCHEDULED
* 1[2]-GOP-123, TURBINE SHUTDOWN - FULL LOAD TO ZERO LOAD
* 1[2]-GOP-201, REACTOR PLANT STARTUP - MODE 2 TO MODE 1
* 1[2]-GOP-203, REACTOR SHUTDOWN
* 1[2]-GOP-302, REACTOR PLANT STARTUP - MODE 3 TO MODE 2
* 1[2]-GOP-303, REACTOR PLANT HEATUP - MODE 3 < 1750 TO MODE 3 > 1750
* 1[2]-GOP-365, REFUELING SEQUENCING GUIDELINES
* 1[2]-GOP-403, REACTOR PLANT HEATUP - MODE 4 TO MODE 3
* 1[2]-GOP-502, DATA SHEETS REQUIRED FOR HEATUP
* 1[2]-GOP-504, REACTOR PLANT HEATUP - MODE 5 TO MODE 4
* 1[2]-AOP-22.01, RAPID DOWNPOWER Page 246 of 246

L-2019-111 Attachment 4 St. Lucie Plant NEI 99-01 Revision 6 EAL Comparison Matrix

EAL Comparison Matrix Table of Contents Section                                                                                                                                                     Page Introduction --------------------------------------------------------------------------------------------------------------------------------------------------------- 1 Comparison Matrix Format ------------------------------------------------------------------------------------------------------------------------------------- 1 EAL Wording ------------------------------------------------------------------------------------------------------------------------------------------------------- 1 EAL Emphasis Techniques ------------------------------------------------------------------------------------------------------------------------------------- 1 Global Differences ------------------------------------------------------------------------------------------------------------------------------------------------ 2 Differences and Deviations ------------------------------------------------------------------------------------------------------------------------------------- 3 Category R - Abnormal Rad Levels / Rad Effluents --------------------------------------------------------------------------------------------------- 13 Category C - Cold Shutdown / Refueling System Malfunction -------------------------------------------------------------------------------------- 31 Category D - Permanently Defueled Station Malfunction -------------------------------------------------------------------------------------------- 51 Category E - Events Related to Independent Spent Fuel Storage Installations ---------------------------------------------------------------- 53 Category F - Fission Product Barrier Degradation ----------------------------------------------------------------------------------------------------- 55 Category H - Hazards and Other Conditions Affecting Plant Safety ------------------------------------------------------------------------------- 68 Category S - System Malfunction -------------------------------------------------------------------------------------------------------------------------- 88 Table 1 - PSL EAL Categories/Subcategories ------------------------------------------------------------------------------------------------------------ 5 Table 2 - NEI / PSL EAL Identification Cross-Reference ----------------------------------------------------------------------------------------------- 6 Table 3 - Summary of Deviations -------------------------------------------------------------------------------------------------------------------------- 11 i of i

EAL Comparison Matrix Introduction                                                                              To assist the Emergency Coordinator (EC), the PSL EALs have been written This document provides a line-by-line comparison of the Initiating Conditions             in a clear and concise style (to the extent that the differences from the NEI (ICs), Mode Applicability and Emergency Action Levels (EALs) in NEI 99-01                 EAL wording could be reasonably documented and justified). This supports Rev. 6 Final, Development of Emergency Action Levels for Non-Passive                      timely and accurate classification in the tense atmosphere of an emergency Reactors, ADAMS Accession Number ML12326A805, and St. Lucie Plant                         event. The EAL differences introduced to reduce reading burden comprise (PSL) ICs, MODE Applicability and EALs. This document provides a means                    almost all of the differences justified in this document.
of assessing PSL differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of PSL EAL bases and lists of                     EAL Emphasis Techniques source document references are given in the EAL Technical Bases                           Due to the width of the table columns and table formatting constraints in this Document. It is, therefore, advisable to reference the EAL Technical Bases                document, line breaks and indentation may differ slightly from the Document for background information while using this document.                            appearance of comparable wording in the source documents. NEI 99-01 Revision 6 is the source document for the NEI EALs; the PSL EAL Technical Comparison Matrix Format                                                                  Bases Document for the PSL EALs.
The ICs and EALs discussed in this document are grouped according to NEI                  Development of the PSL IC/EAL wording has attempted to minimize 99-01 Recognition Categories. Within each Recognition Category, the ICs                   inconsistencies and apply sound human factors principles. As a result, and EALs are listed in tabular format according to the order in which they are            differences occur between NEI and PSL ICs/EALs for these reasons alone.
given in NEI 99-01. Generally, each row of the comparison matrix provides                 When such difference may infer a technical difference in the associated NEI the following information:                                                                IC/EAL, the difference is identified and a justification provided.
The print and paragraph formatting conventions summarized below guide
* NEI EAL/IC identifier presentation of the PSL EALs in accordance with the EAL writing criteria.
* NEI EAL/IC wording                                                               Space restrictions in the EAL table of this document sometimes override this
* PSL EAL/IC identifier                                                            criteria in cases when following the criteria would introduce undesirable complications in the EAL layout.
* PSL EAL/IC wording
* Upper case-bold underline print is used for the logic terms AND, OR
* Description of any differences or deviations                                               and EITHER.
* Bold print is also used for certain logic terms, negative terms (not, EAL Wording                                                                                         cannot, etc.), any, all.
In Section 4.1, NEI recommends the following: The guidance in NEI 99-01 is
* Upper case print is reserved for defined terms, acronyms, system not intended to be applied to plants as-is; however, developers should                            abbreviations, logic terms (and, or, etc. when not used as a attempt to keep their site-specific schemes as close to the generic guidance                        conjunction), annunciator window engravings.
as possible. The goal is to meet the intent of the generic Initiating Conditions
* Three or more items in a list are normally introduced with Any of the (ICs) and Emergency Action Levels (EALs) within the context of site-specific                        following... or All of the following... Items of the list begin with characteristics - locale, plant design, operating features, terminology, etc.                       bullets when a priority or sequence is not inferred.
Meeting this goal will result in a shorter and less cumbersome NRC review and approval process, closer alignment with the schemes of other nuclear
* The use of and/or logic within the same EAL has been avoided power plant sites and better positioning to adopt future industry-wide scheme                       when possible. When such logic cannot be avoided, indentation and enhancements                                                                                       separation of subordinate contingent phrases is employed.
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EAL Comparison Matrix
* NEI 99-01 defines the thresholds requiring emergency Global Differences                                                                         classification (example EALs) and assigns them to ICs which, in turn, are grouped in Recognition Categories. PSL endeavors to The differences listed below generally apply throughout the set of EALs and                optimize the NEI EAL organization and identification scheme to are not repeated in the Justification sections of this document. The global                enhance usability of the plant-specific EAL set. To this end, the differences do not decrease the effectiveness of the intent of NEI 99-01.                  PSL IC/EAL scheme includes the following features:
: 1. The NEI phrase Notification of Unusual Event has been changed to                    a. Division of the NEI EAL set into three groups:
Unusual Event or abbreviated "UE" to reduce EAL-user reading burden.                                                                                  o EALs applicable under all plant operating conditions -
This group would be reviewed by the EAL-user any
: 2. The title Emergency Director is replaced with the PSL-specific title time emergency classification is considered.
Emergency Coordinator (EC)
: 3. NEI 99-01 IC Example EALs are implemented in separate plant o EALs applicable only under hot operating conditions -
This group would only be reviewed by the EAL-user EALs to improve clarity and readability. For example, NEI lists all IC when the plant is in Power Operation, Startup, Hot HU3 Example EALs under one IC. The corresponding PSL EALs Standby or Hot Shutdown mode.
appear as unique EALs (e.g., HU3.1 through HU3.4).
: 4. Operational Condition (MODE) applicability identifiers o EALs applicable only under cold operating conditions -
This group would only be reviewed by the EAL-user (numbers/letter) modify the NEI 99-01 mode applicability names as when the plant is in Cold Shutdown, Refueling or follows: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Defueled mode.
Shutdown, 5 - Cold Shutdown, 6 - Refueling, DEF - Defueled. NEI 99-01 defines Defueled as follows: All reactor fuel removed from                         The purpose of the groups is to avoid review of hot RPV. (Full core off load during refueling or extended outage).                           condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a
: 5. NEI 99-01 uses the terms greater than, less than, greater than or hot condition. This approach significantly minimizes the equal to, etc. in the wording of some example EALs. For consistency total number of EALs that must be reviewed by the EAL-and to reduce EAL-user reading burden, PSL has adopted use of user for a given plant condition, reduces EAL-user boolean symbols in place of the NEI 99-01 text modifiers within the reading burden and, thereby, speeds identification of the EAL wording.
EAL that applies to the emergency.
: 6. min. is the standard abbreviation for minutes and is used to
: b. Within each of the above three groups, assignment of reduce EAL user reading burden.
EALs to categories/subcategories - Category and
: 7. IC/EAL identification:                                                                     subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.
* NEI Recognition Category A Abnormal Radiation Levels/
Subcategories are used as necessary to further divide the Radiological Effluents has been changed to Category R EALs of a category into logical sets of possible Abnormal Rad Levels / Rad Effluents. The designator R is emergency classification thresholds. The PSL EAL more intuitively associated with radiation (rad) or radiological categories/subcategories and their relationship to NEI events. NEI IC designators beginning with A have likewise Recognition Categories are listed in Table 1.
been changed to R.
: c. Unique identification of each EAL - Four characters comprise the EAL identifier as illustrated in Figure 1.
2 of 124

EAL Comparison Matrix o Possible classification upgrade - The Figure 1 - EAL Identifier                                                                                category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs EAL Identifier                                                                                         that may become active if plant conditions worsen.
XXX.X Category (R, H, E, S, F, C)         Sequential number within subcategory/classification Table 2 lists the PSL ICs and EALs that correspond to the Emergency classification (G, S, A, U)       Subcategory number (1 if no subcategory)
NEI ICs/Example EALs when the above EAL/IC organization and identification scheme is implemented.
The first character is a letter associated with the category in which the EAL is located. The second character is a                                         Differences and Deviations letter associated with the emergency classification level In accordance NRC Regulatory Issue Summary (RIS) 2003-18 Use of (G for General Emergency, S for Site Area Emergency, A Nuclear Energy Institute (NEI) 99-01, Methodology for Development of for Alert, and U for Unusual Event). The third character is Emergency Action Levels Supplements 1 and 2, a difference is an EAL a number associated with one or more subcategories change in which the basis scheme guidance differs in wording but agrees in within a given category. Subcategories are sequentially meaning and intent, such that classification of an event would be the same, numbered beginning with the number 1. If a category whether using the basis scheme guidance or the PSL EAL. A deviation is an does not have a subcategory, this character is assigned EAL change in which the basis scheme guidance differs in wording and is the number 1. The fourth character is a number altered in meaning or intent, such that classification of the event could be preceded by a period for each EAL within a subcategory.
different between the basis scheme guidance and the PSL proposed EAL.
EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the                                       Administrative changes that do not actually change the textual content are number 1.                                                                                    neither differences nor deviations. Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a The EAL identifier is designed to fulfill the following deviation.
objectives:
The following are examples of differences:
o Uniqueness - The EAL identifier ensures that there can be no confusion over which EAL is driving the
* Choosing the applicable EAL based upon plant type (i.e., BWR vs.
need for emergency classification.                                                         PWR).
o Speed in locating the EAL of concern - When the
* Using a numbering scheme other than that provided in NEI 99-01 EALs are displayed in a matrix format, knowledge                                           that does not change the intent of the overall scheme.
of the EAL identifier alone can lead the EAL-user to
* Where the NEI 99-01 guidance specifically provides an option to not the location of the EAL within the classification include an EAL if equipment for the EAL does not exist at PSL (e.g.,
matrix. The identifier conveys the category, automatic real-time dose assessment capability).
subcategory and classification level. This assists ERO responders (who may not be in the same
* Pulling information from the bases section up to the actual EAL that facility as the EC) to find the EAL of concern in a                                        does not change the intent of the EAL.
timely manner without the need for a word
* Choosing to state ALL Operating Modes are applicable instead of description of the classification threshold.                                               stating N/A, or listing each mode individually under the Abnormal 3 of 124

EAL Comparison Matrix Rad Level/Radiological Effluent and Hazard and Other Conditions                         all users to have a standard set of defined terms as defined in NEI Affecting Plant Safety sections.                                                        99-01.
* Using synonymous wording (e.g., greater than or equal to vs. at or
* Any change to the IC and/or EAL, and/or basis wording as stated in above, less than or equal vs. at or below, greater than or less than                    NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC vs. above or below, etc.)                                                               and/or EAL:
* Adding PSL equipment/instrument identification and/or noun names                             o   Does not classify at the classification level consistent with to EALs.                                                                                         NEI 99-01.
* Combining like ICs that are exactly the same but have different                              o   Is not logically integrated with other EALs in the EAL operating modes as long as the intent of each IC is maintained and                               scheme.
the overall progression of the EAL scheme is not affected.                                   o   Results in an incomplete EAL scheme (i.e., does not classify
* Any change to the IC and/or EAL, and/or basis wording, as stated in                              all potential emergency conditions).
NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e.,          The Difference/Deviation Justification column in the remaining sections of the IC and/or EAL continues to:                                                this document identify each difference between the NEI 99-01 IC/EAL o   Classify at the correct classification level.                         wording and the PSL IC/EAL wording. An explanation that justifies the reason for each difference is then provided. If the difference is determined to o   Logically integrate with other EALs in the EAL scheme.                be a deviation, a statement is made to that affect and explanation is given o   Ensure that the resulting EAL scheme is complete (i.e.,               that states why classification may be different from the NEI 99-01 IC/EAL and classifies all potential emergency conditions).                       the reason for its acceptability. In all cases, however, the differences and The following are examples of deviations:                                               deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of PSL EAL deviations from NEI 99-01 is given in Table 3.
* Use of altered mode applicability.
* Altering key words or time limits.
* Changing words of physical reference (protected area, safety-related equipment, etc.).
* Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier ICs.
* Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa.
* Not using NEI 99-01 definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-
: 01. Differences due to plant types are permissible (BWR or PWR).
Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained. Use of the wording provided in NEI 99-01 is encouraged since the intent is for 4 of 124

EAL Comparison Matrix Table 1 - PSL EAL Categories/Subcategories PSL EALs                                                           NEI Category                                          Subcategory                       Recognition Category Group: Any Operating Mode:
1 - Radiological Effluent                    Abnormal Rad Levels/Radiological Effluent 2 - Irradiated Fuel Event R - Abnormal Rad Levels/Rad Effluent                                                        ICs/EALs 3 - Area Radiation Levels H - Hazards and Other Conditions Affecting     1 - Security                                 Hazards and Other Conditions Affecting Plant Safety                               2 - Seismic Event Plant Safety ICs/EALs 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - EC Judgment E - ISFSI                                      1 - Confinement Boundary                     ISFSI ICs/EALs Group: Hot Conditions:
1 - Loss of Emergency AC Power               System Malfunction ICs/EALs 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity S - System Malfunction                         5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier                    None                                         Fission Product Barrier ICs/EALs Group: Cold Conditions:
1 - RCS Level                                Cold Shutdown./ Refueling System 2 - Loss of Emergency AC Power               Malfunction ICs/EALs C - Cold Shutdown/Refueling System             3 - RCS Temperature Malfunction                                4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems 5 of 124

EAL Comparison Matrix Table 2 - NEI / PSL EAL Identification Cross-Reference NEI                                                          PSL Example IC                                             Category and Subcategory            EAL EAL AU1         1    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RU1.1 AU1         2    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RU1.2 AU1         3    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RU1.3 AU2         1    R - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event RU2.1 AA1         1    A - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RA1.1 AA1         2    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RA1.2 AA1         3    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RA1.3 AA1         4    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RA1.4 AA2         1    R - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event RA2.1 AA2         2    R - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event RA2.2 AA2         3    R - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event RA2.3 AA3         1    R - Abnormal Rad Levels / Rad Effluent, 3 - Area Radiation Levels RA3.1 AA3         2    R - Abnormal Rad Levels / Rad Effluent, 3 - Area Radiation Levels RA3.2 AS1         1    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RS1.1 AS1         2    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RS1.2 AS1         3    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent RS1.3 6 of 124

EAL Comparison Matrix NEI                                                            PSL Example IC                                               Category and Subcategory                                  EAL EAL AS2           1    R - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event                       RS2.1 AG1           1    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent                       RG1.1 AG1           2    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent                       RG1.2 AG1           3    R - Abnormal Rad Levels / Rad Effluent, 1 - Radiological Effluent                       RG1.3 AG2           1    R - Abnormal Rad Levels / Rad Effluent, 2 - Irradiated Fuel Event                       RG2.1 CU1           1    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                CU1.1 CU1           2    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                CU1.2 CU2           1    C - Cold SD/ Refueling System Malfunction, 2 - Loss of AC Power                         CU2.1 CU3           1    C - Cold SD/ Refueling System Malfunction, 3 - RCS Temperature                          CU3.1 CU3           2    C - Cold SD/ Refueling System Malfunction, 3 - RCS Temperature                          CU3.2 CU4           1    C - Cold SD/ Refueling System Malfunction, 4 - Loss of DC Power                         CU4.1 CU5        1, 2, 3 C - Cold SD/ Refueling System Malfunction, 5 - Loss of Communications                   CU5.1 CA1           1    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                CA1.1 CA1           2    C - Cold SD/ Refueling System Malfunction, 1 - RCSV Level                               CA1.2 CA2           1    C - Cold SD/ Refueling System Malfunction, 1 - Loss of AC Power                         CA2.1 CA3         1, 2   C - Cold SD/ Refueling System Malfunction, 3 - RCS Temperature                          CA3.1 CA6           1    C - Cold SD/ Refueling System Malfunction, 6 - Hazardous Event Affecting Safety Systems CA6.1 CS1           1    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                CS1.1 7 of 124

EAL Comparison Matrix NEI                                                           PSL Example IC                                               Category and Subcategory                                    EAL EAL CS1           2    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                   N/A CS1           3    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                  CS1.2 CG1            1    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                  CG1.1 CG1            2    C - Cold SD/ Refueling System Malfunction, 1 - RCS Level                                  CG1.2 E-HU1           1    E - ISFSI, 1 - Confinement Boundary                                                       EU1.1 FA1           1    F - Fission Product Barrier                                                               FA1.1 FS1           1    F - Fission Product Barrier                                                               FS1.1 FG1           1    F - Fission Product Barrier                                                               FG1.1 HU1        1, 2, 3 H - Hazards and Other Conditions Affecting Plant Safety, 1 - Security                     HU1.1 HU2           1    H - Hazards and Other Conditions Affecting Plant Safety, 2 - Seismic Event                HU2.1 HU3           1    H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technology Hazard HU3.1 HU3           2    H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technology Hazard HU3.2 HU3           3    H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technology Hazard HU3.3 HU3           4    H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technology Hazard HU3.4 HU3           5    H - Hazards and Other Conditions Affecting Plant Safety, 3 - Natural or Technology Hazard  N/A HU4           1    H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion            HU4.1 HU4           2    H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion            HU4.2 HU4           3    H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion            HU4.3 8 of 124

EAL Comparison Matrix NEI                                                           PSL Example IC                                                Category and Subcategory                              EAL EAL HU4           4    H - Hazards and Other Conditions Affecting Plant Safety, 4 - Fire or Explosion       HU4.4 HU7           1    H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment                HU7.1 HA1         1, 2   H - Hazards and Other Conditions Affecting Plant Safety, 1 - Security                HA1.1 HA5           1    H - Hazards and Other Conditions Affecting Plant Safety, 5 - Hazardous Gases         HA5.1 HA6           1    H - Hazards and Other Conditions Affecting Plant Safety, 6 - Control Room Evacuation HA6.1 HA7           1    H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment                HA7.1 HS1           1    H - Hazards and Other Conditions Affecting Plant Safety, 1 - Security                HS1.1 HS6           1    H - Hazards and Other Conditions Affecting Plant Safety, 6 - Control Room Evacuation HS6.1 HS7           1    H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment                HS7.1 HG1            1    N/A                                                                                   N/A HG7            1    H - Hazards and Other Conditions Affecting Plant Safety, 7 - Judgment                HG7.1 SU1           1    S - System Malfunction, 1 - Loss of AC Power                                         SU1.1 SU2           1    S - System Malfunction, 3 - Loss of Control Room Indications                         SU3.1 SU3           1    N/A                                                                                   N/A SU3           2    S - System Malfunction, 4 - RCS Activity                                             SU4.1 SU4        1, 2, 3 S - System Malfunction, 5 - RCS Leakage                                              SU5.1 SU5           1    S - System Malfunction, 6 - RPS Failure                                              SU6.1 SU5           2    S - System Malfunction, 6 - RPS Failure                                              SU6.2 9 of 124

EAL Comparison Matrix NEI                                                           PSL Example IC                                                Category and Subcategory      EAL EAL SU6        1, 2, 3 S - System Malfunction, 7 -Loss of Communications            SU7.1 SU7         1, 2   S - System Malfunction, 8 - Containment Failure              SU8.1 SA1           1    S - System Malfunction, 1 - Loss of AC Power                 SA1.1 SA2           1    S - System Malfunction, 3 - Loss of Control Room Indications SA3.1 SA5           1    S - System Malfunction, 6 - RPS Failure                      SA6.1 SA9           1    S - Hazardous Event Affecting Safety Systems                 SA9.1 SS1           1    S - System Malfunction, 1 - Loss of AC Power                 SS1.1 SS5           1    S - System Malfunction, 6 - RPS Failure                      SS6.1 SS8           1    S - System Malfunction, 2 - Loss of DC Power                 SS2.1 SG1           1    S - System Malfunction, 1 - Loss of AC Power                 SG1.1 SG8           2    S - System Malfunction, 2 - Loss of DC Power                 SG2.1 10 of 124

EAL Comparison Matrix Table 3 - Summary of Deviations NEI PSL Description EAL IC        Example EAL HG1            1       N/A      IC HG1 and associated example EAL is not implemented in the PSL scheme.
There are several other ICs that are redundant with this IC, and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA                                                051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified because:
: 1. Hostile Action in the Protected Area is bounded by ICs HS1 and HS7. Hostile Action resulting in a loss of physical control is bound by EAL HG7, as well as any event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA) Protective Action Guides (PAGs).
: a. If, for whatever reason, the Control Room must be evacuated, and control of safety functions (e.g., reactivity control, core cooling, and RCS heat removal) cannot be reestablished, then IC HS6 would apply, as well as IC HS7 if desired by the EAL decision-maker.
: b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG7.
: c. From a Hostile Action perspective, ICs HS1, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
: d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
: 2. Any event which causes a loss of spent fuel pool level will be bounded by ICs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.
: a. An event that leads to a radiological release will be bounded by ICs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of 11 of 124

EAL Comparison Matrix NEI PSL Description EAL IC        Example EAL EPA PAGs will be bounded by EALs AG1 and HG7, thus making this part of HG1 redundant and unnecessary.
ICs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG7 have been implemented consistent with NEI 99-01 Revision 6 and thus HG1 is adequately bounded as described above.
This exclusion of the generic HG1 guidance is a deviation from the NEI 99-01 Revision generic guidance but is deemed acceptable consistent with endorsed NRC EP FAQ 2015-013.
HS6            1       HS6.1 Deleted defueled mode applicability. Control of the cited safety functions are not critical for a defueled reactor as there is no energy source in the reactor vessel or RCS.
The Mode applicability for the reactivity control safety function has been limited to Modes 1, 2, and 3 (hot operating conditions). In the cold operating modes adequate shutdown margin exists under all conditions.
This revised mode applicability is a deviation from the NEI 99-01 Revision 6 HS6 generic guidance but is deemed acceptable consistent with endorsed NRC EP FAQ 2015-014.
CA6            1       CA6.1 The proposed PSL CA6.1 and SA9.1 wording is intended to ensure that an Alert should be declared only when actual or potential performance issues with SA9            1       SA9.1 SAFETY SYSTEMS have occurred as a result of a hazardous event. The occurrence of a hazardous event will result in an Unusual Event classification at a minimum. In order to warrant escalation to the Alert classification, the hazardous event should cause indications of degraded performance to one train of a SAFETY SYSTEM with either indications of degraded performance on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second SAFETY SYSTEM train, such that the operability or reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable, or out of service, prior to the event occurring. As such, the proposed EALs will reduce the potential of declaring an Alert when events are in progress that do not involve an actual or potential substantial degradation of the level of safety of the plant, i.e., does not cause significant concern with shutting down or cooling down the 12 of 124

EAL Comparison Matrix NEI PSL Description EAL IC        Example EAL plant.
EALs CA6.1 and SA9.1 do not directly escalate to a Site Area Emergency or a General Emergency due to a hazardous event. The Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency.
The EALs and the Basis sections have been revised to ensure potential escalations from an Unusual Event to an Alert, due to a hazardous event, is appropriate as the concern with these EALs is: (1) a hazardous event has occurred, (2) one SAFETY SYSTEM train is having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or the VISIBLE DAMAGE is enough to be concerned that the second SAFETY SYSTEM train may have operability or reliability issues.
The definition for VISIBLE DAMAGE has been revised to reflect the fact that the EALs are based upon SAFETY SYSTEM trains rather than individual components or structures.
Note 9 has been added to CA6.1 and SA9.1 as it meets the intent of the EALs, is consistent with other EALs (e.g., EAL HA5.1 which was previously endorsed by the NRC), and ensures that declared emergencies are based upon unplanned events with the potential to pose a radiological risk to the public.
Note 10 has been added to CA6.1 and SA9.1 to help reinforce and succinctly capture the more detailed information from the revised basis section related to when conditions would require the declaration of an Alert.
CA6.1 and SA9.1 are consistent with NRC FAQ 2016-002 requiring degraded performance or visible damage to more than one safety system train caused by the specified events.
This revised wording is a deviation from the NEI 99-01 Revision 6 CA6 and SA9 generic wording and bases but is deemed acceptable consistent with endorsed NRC EP FAQ 2016-002.
SG1            1       SG1.1 The proposed PSL SG1.1 omits the Station Blackout (SBO) coping time threshold. As proposed, the General Emergency classification would be based on a loss of all onsite and offsite AC power to the emergency buses with indications 13 of 124

EAL Comparison Matrix NEI PSL Description EAL IC         Example EAL of degraded core cooling. The PSL SBO analysis and derived coping time was determined in accordance with 10CFR50.63 and Regulatory Guide 1.155. This analysis does not take credit for plant capabilities in place to mitigate the effects of an extended loss of AC power (ELAP). These capabilities were developed and implemented to meet the requirements of NRC Orders EA-12-049 and EA                                           051, and pending regulations in 10 CFR 50.155 (per SECY-16-0142).
In accordance with plant EOPs [1(2)-EOP-10], operators will declare an ELAP within 60 min. of the loss of all AC power to the emergency buses and direct implementation of FLEX Support Guidelines, including the deployment of dedicated portable equipment and performance of DC load shedding. Even if no AC emergency bus is energized, these actions will maintain or restore core cooling, containment, and spent fuel pool cooling capabilities indefinitely. Therefore, the underlying basis for the generic EAL coping time statement, that power must be restored to an AC emergency bus within a fixed amount of time to avoid a severe challenge to one or more fission product barriers, is not valid for PSL.
Additionally, the omission of the SBO coping time threshold does not remove the attribute of a likely General Emergency declaration prior to meeting the IC FG1 thresholds for ELAP events in which the RCS barrier has not been lost.
This revised wording is a deviation from the NEI 99-01 Revision 6 SG1 generic wording and bases but is deemed appropriate and acceptable.
14 of 124

EAL Comparison Matrix Category A Abnormal Rad Levels / Radiological Effluent 15 of 124

EAL Comparison Matrix NEI IC Wording and Mode          PSL           PSL IC Wording and Mode NEI IC#                                                                                                        Difference/Deviation Justification Applicability           IC#(s)                 Applicability AU1    Release of gaseous or liquid          RU1    Release of gaseous or liquid             The PSL ODCM is the site-specific effluent release radioactivity greater than 2 times           radioactivity greater than 2 times the   controlling document.
the (site-specific effluent release          ODCM limits for 60 minutes or longer controlling document) limits for             MODE: All 60 minutes or longer.
MODE: All NEI Ex.                                     PSL EAL NEI Example EAL Wording                            PSL EAL Wording                             Difference/Deviation Justification EAL #                                          #
1      Reading on ANY effluent                    Reading on any Table 1[2]R-1 effluent   The NEI phrase effluent radiation monitor greater than 2 radiation monitor greater than 2           radiation monitor > column "UE" for     times the (site-specific effluent release controlling times the (site-specific effluent           60 min. (Notes 1, 2, 3)               document) has been replaced with any Table 1[2]R-1 release controlling document)                                                      effluent radiation monitor > column "UE".
limits for 60 minutes or longer:
UE thresholds for all PSL continuously monitored gaseous (site-specific monitor list and      RU1.1                                          and liquid release pathways are listed in Table 1[2]R-1 to threshold values corresponding                                                      consolidate the information in a single location and, thereby, to 2 times the controlling                                                          simplify identification of the thresholds by the EAL user. The document limits)                                                                    values shown in Table 1[2]R-1 column UE, consistent with the NEI bases, represent two times the ODCM release limits for gaseous and liquid releases.
2     Reading on ANY effluent                     Reading on any effluent radiation       None radiation monitor greater than 2            monitors > 2 x the alarm setpoint times the alarm setpoint                    established by a current radioactivity RU1.2  discharge permit for  60 min. (Notes established by a current radioactivity discharge permit for          1, 2, 3) 60 minutes or longer.
3     Sample analysis for a gaseous               Sample analysis for a gaseous or        The PSL ODCM is the site-specific effluent release or liquid release indicates a               liquid release indicates a              controlling document.
concentration or release rate        RU1.3  concentration or release rate > 2 x greater than 2 times the (site-             ODCM limits for  60 min. (Notes 1, 2) specific effluent release 16 of 124

EAL Comparison Matrix NEI Ex.                                    PSL EAL NEI Example EAL Wording                         PSL EAL Wording                            Difference/Deviation Justification EAL #                                         #
controlling document) limits for 60 minutes or longer.
Notes     The Emergency Director           N/A   Note 1: The EC should declare the    The classification timeliness note has been standardized should declare the Unusual                     event promptly upon          across the PSL EAL scheme by referencing the "time limit" Event promptly upon                            determining that the time    specified within the EAL wording.
determining that 60 minutes                    limit has been exceeded, or has been exceeded, or will                     will likely be exceeded.
likely be exceeded.
Note 2: If an ongoing release is If an ongoing release is                       detected and the release     The classification timeliness note has been standardized detected and the release                       start time is unknown,       across the PSL EAL scheme by referencing the "time limit" start time is unknown,                         assume that the release      specified within the EAL wording.
assume that the release                        duration has exceeded the duration has exceeded 60                       specified time limit.
minutes.
Note 3: If the effluent flow past an If the effluent flow past an                   effluent monitor is known to effluent monitor is known to                   have stopped due to actions  None have stopped due to actions                    to isolate the release path, to isolate the release path,                   then the effluent monitor then the effluent monitor                      reading is no longer VALID reading is no longer valid for                 for classification purposes.
classification purposes.
17 of 124

EAL Comparison Matrix Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds Release Point & Monitor             GE               SAE               Alert              UE Plant Vent 1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc   2.92E-03 &#xb5;Ci/cc RSC 26-1 (L/H)
Gaseous ECCS Vent RSC 26-2 (L/H) (A) 7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RSC 26-3 (L/H) (B)
FHB Vent 2.23E+03 &#xb5;Ci/cc   2.23E+02 &#xb5;Ci/cc   2.23E+01 &#xb5;Ci/cc   9.65E-03 &#xb5;Ci/cc RSC 26-4 (L/H)
Liquid Waste Discharge N/A               N/A               N/A          1.33E-03 &#xb5;Ci/ml RSC-26-43 Liquid    SG Blowdown A" SG RE-26-44-1A             N/A               N/A               N/A          2.44E+04 cpm B" SG RE-26-45-1B             N/A               N/A               N/A          5.81E+04 cpm 18 of 124

EAL Comparison Matrix Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds Release Point & Monitor               GE               SAE               Alert              UE Plant Vent RIM-26-90 (621/622/623) 2.43E-03 &#xb5;Ci/cc
                                   - Normal or post SIAS      1.55E+02 &#xb5;Ci/cc   1.55E+01 &#xb5;Ci/cc   1.55E+00 &#xb5;Ci/cc Gaseous
                                   - w/ FHB diverted to PV    1.78E+03 &#xb5;Ci/cc   1.78E+02 &#xb5;Ci/cc   1.78E+01 &#xb5;Ci/cc ECCS Vent RIM-26-69 (601/602/603) (A)   7.84E+01 &#xb5;Ci/cc   7.84E+00 &#xb5;Ci/cc   7.84E-01 &#xb5;Ci/cc        N/A RIM-26-70 (611/612/613) (B)
SG Blowdown Liquid      RM-26-5-2A                         N/A               N/A               N/A          5.81E-04 &#xb5;Ci/ml RM-26-6-2B 19 of 124

EAL Comparison Matrix NEI IC Wording and Mode         PSL          PSL IC Wording and Mode NEI IC#                                                                                                    Difference/Deviation Justification Applicability          IC#(s)                Applicability AU2    UNPLANNED loss of water level       RU2    UNPLANNED loss of water level          None above irradiated fuel.                     above irradiated fuel MODE: All                                  MODE: All NEI Ex.                                     PSL NEI Example EAL Wording                          PSL EAL Wording                             Difference/Deviation Justification EAL #                                      EAL #
1     a. UNPLANNED water level           RU2.1  UNPLANNED water level drop in the      Site-specific level indications incorporated.
drop in the REFUELING                 REFUELING PATHWAY as indicated Site-specific area radiation monitors tabularized in Tables PATHWAY as indicated by               by any of the following:
1[2]R-2.
ANY of the following:
(site-specific level
* SFP or reactor cavity level Added the word corresponding to reinforce the cause indications).                            instrument (water level decrease) and effect (area radiation levels) intent AND
* SFP or reactor cavity low level    of this EAL.
: b. UNPLANNED rise in area                     alarm radiation levels as indicated by ANY of the following
* Report of dropping level in SFP or radiation monitors.                      reactor cavity (site-specific list of area             AND radiation monitors)                  UNPLANNED rise in corresponding area radiation levels as indicated by any Table 1[2]R-2 radiation monitors 20 of 124

EAL Comparison Matrix Table 1R-2 Unit 1 Refueling Area Radiation Monitors ARM #         Channel #                  Description RE26-4-1             4         Containment CIS Monitor RE26-6-1             6         Containment CIS Monitor RE26-7-1             7         Spent Fuel Pool RE26-8-1             8         Refueling Canal RE26-36-1           36         Operating Deck Table 2R-2 Unit 2 Refueling Area Radiation Monitors ARM #      RMCS Channel                  Description RIM-26-2         GAG-002       CCW RIM-26-4         GAG-004       Containment CIS Monitor B RIM-26-6         GAG-006       Containment CIS Monitor D RIM-26-7         GAG-007       Spent Fuel Pool RIM-26-8         GAG-008       Spent Fuel Pool RIM-26-9         GAG-009       Spent Fuel Pool RIM-26-10        GAG-010       Spent Fuel Pool RIM-26-11        GAG-011       Spent Fuel Pool RIM-26-12        GAG-012       Spent Fuel Pool 21 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                                PSL IC Wording                              Difference/Deviation Justification IC#(s)
AA1    Release of gaseous or liquid         RA1    Release of gaseous or liquid             None radioactivity resulting in offsite          radioactivity resulting in offsite dose dose greater than 10 mrem TEDE              greater than 10 mrem TEDE or 50 or 50 mrem thyroid CDE.                     mrem thyroid CDE MODE: All                                   MODE: All NEI Ex.                                     PSL NEI Example EAL Wording                           PSL EAL Wording                               Difference/Deviation Justification EAL #                                      EAL #
1     Reading on ANY of the following    RA1.1   Reading on any Table 1[2]R-1 effluent    The PSL radiation monitors that detect radioactivity effluent radiation monitors greater than            radiation monitor > column "ALERT"       release to the environment are listed in Table 1[2]R-1. UE, the reading shown for 15 minutes           for  15 min. (Notes 1, 2, 3, 4)         Alert, SAE and GE thresholds for all PSL continuously or longer:
monitored gaseous and liquid release pathways are listed in Table 1[2]R-1 to consolidate the information in a single (site-specific monitor list and location and, thereby, simplify identification of the thresholds threshold values) by the EAL-user.
2     Dose assessment using actual       RA1.2   Dose assessment using actual             The site boundary is the site-specific receptor point.
meteorology indicates doses                meteorology indicates doses > 10 greater than 10 mrem TEDE or               mrem TEDE or 50 mrem thyroid CDE 50 mrem thyroid CDE at or                  at or beyond the SITE BOUNDARY beyond (site-specific dose                 (Note 4) receptor point).
3     Analysis of a liquid effluent      RA1.3   Analysis of a liquid effluent sample     The site boundary is the site-specific receptor point.
sample indicates a concentration           indicates a concentration or release or release rate that would result          rate that would result in doses > 10 in doses greater than 10 mrem              mrem TEDE or 50 mrem thyroid CDE TEDE or 50 mrem thyroid CDE at             at or beyond the SITE BOUNDARY for or beyond (site-specific dose              60 min. of exposure (Notes 1, 2) receptor point) for one hour of exposure.
22 of 124

EAL Comparison Matrix 4     Field survey results indicate      RA1.4 Field survey results indicate EITHER       The site boundary is the site-specific field survey receptor EITHER of the following at or of the following at or beyond the SITE     point.
beyond (site-specific dose BOUNDARY:
receptor point):
Closed window dose rates                 Closed window dose rates > 10 greater than 10 mR/hr                    mR/hr expected to continue for expected to continue for 60              60 min.
minutes or longer.
Analyses of field survey samples Analyses of field survey indicate thyroid CDE > 50 mrem samples indicate thyroid CDE greater than 50 mrem                 for 60 min. of inhalation.
for one hour of inhalation.         (Notes 1, 2)
Notes      The Emergency Director          N/A  Note 1:     The EC should declare the      The classification timeliness note has been standardized should declare the Alert                         event promptly upon            across the PSL EAL scheme by referencing the "time limit" promptly upon determining                        determining that the time      specified within the EAL wording.
that the applicable time has                     limit has been exceeded, or been exceeded, or will likely                    will likely be exceeded.
be exceeded.
Note 2:     If an ongoing release is If an ongoing release is                         detected and the release       The classification timeliness note has been standardized detected and the release                         start time is unknown,         across the PSL EAL scheme by referencing the "time limit" start time is unknown,                           assume that the release        specified within the EAL wording.
assume that the release                          duration has exceeded the duration has exceeded 15                         specified time limit.
minutes.
Note 3:     If the effluent flow past an If the effluent flow past an                     effluent monitor is known to effluent monitor is known to                     have stopped due to actions    None have stopped due to actions                      to isolate the release path, to isolate the release path,                     then the effluent monitor then the effluent monitor                        reading is no longer valid for reading is no longer valid for                   classification purposes s.
classification purposes.
Note 4      The pre-calculated effluent monitor values presented in    Incorporated site-specific EAL numbers associated with The pre-calculated effluent                      EALs RA1.1, RS1.1 and          generic EAL#1.
monitor values presented in                      RG1.1 should be used for EAL #1 should be used for                        emergency classification emergency classification                         assessments until the assessments until the results                    results from a dose 23 of 124

EAL Comparison Matrix from a dose assessment       assessment using actual using actual meteorology are meteorology are available.
available.
24 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                                 PSL IC Wording                                  Difference/Deviation Justification IC#(s)
AA2     Significant lowering of water       RA2   Significant lowering of water level            None level above, or damage to,                above, or damage to, irradiated fuel irradiated fuel.                          MODE: All MODE: All NEI Ex.                                       PSL NEI Example EAL Wording                             PSL EAL Wording                                  Difference/Deviation Justification EAL #                                        EAL #
1     Uncovery of irradiated fuel in the   RA2.1  IMMINENT uncovery of irradiated fuel         Added the term IMMINENT consistent with the generic REFUELING PATHWAY.                          in the REFUELING PATHWAY                     bases.
2      Damage to irradiated fuel           RA2.2 Damage to irradiated fuel resulting in a      Deleted the words from the fuel as that is implied by the resulting in a release of release of radioactivity                      determination that irradiated fuel has been damaged.
radioactivity from the fuel as indicated by ANY of the                         AND                                     Site-specific list of radiation monitors are incorporated.
following radiation monitors:
VALID HIGH alarm on any of the                Tabularized the Unit 1[2] area radiation monitors in Tables following:                                    1[2]R-2.
(site-specific listing of radiation monitors, and the associated
* Any Table 1[2]R-2 area radiation         Valid radiation monitor high alarms specified.
readings, setpoints and/or                       monitors alarms)
* RSC 26-4 [RSC-26-12] Fuel Handling Building Exhaust Vent 3      Lowering of spent fuel pool level   RA2.3  Lowering of spent fuel pool level to 46      For PSL, Level 2 corresponds to 46 ft. 3 in [46 ft. 5 in.] above to (site-specific Level 2 value).          ft. 3 in. [46 ft. 5 in.] (Level 2) on LI the top of the fuel racks in the SFP, on LI-04-1[2]A or LI           [See Developer Notes]                      1[2]A or LI-04-1[2]B                         1[2]B.
25 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                                     PSL IC Wording                            Difference/Deviation Justification IC#(s)
AA3    Radiation levels that impede             RA3   Radiation levels that IMPEDE access       Limited mode applicability of RA3.2 modes specified in Table access to equipment necessary                  to equipment necessary for normal         R-3.
for normal plant operations,                   plant operations, cooldown or cooldown or shutdown                           shutdown MODE: All                                      MODE: All (except RA3.2: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown only)
NEI Ex.                                          PSL NEI Example EAL Wording                                PSL EAL Wording                             Difference/Deviation Justification EAL #                                           EAL #
Dose rate greater than 15 mR/hr 1                                             RA3.1  Dose rates > 15 mR/hr in EITHER of        No other site-specific areas requiring continuous occupancy in ANY of the following areas:
the following:                            exist at PSL.
Control Room Central Alarm Station
* Control Room (other site-specific areas/rooms)
* Central Alarm Station 2      An UNPLANNED event results             RA3.2   An UNPLANNED event results in            The site-specific list of plant rooms or areas with entry-related in radiation levels that prohibit or           radiation levels that prohibit or        mode applicability are tabularized in Table R-3.
impede access to any of the                    IMPEDE access to any Table R-3 following plant rooms or areas:                room or area (Note 5)
(site-specific list of plant rooms or areas with entry-related mode applicability identified)
Note    If the equipment in the listed room or  N/A    Note 5   If the equipment in the listed  None area was already inoperable or out-                     room or area was already of-service before the event                             inoperable or out-of-service occurred, then no emergency                             before the event occurred, then classification is warranted.                            no emergency classification is warranted.
26 of 124

EAL Comparison Matrix N/A    N/A                  N/A       Note 15: The Intake Structure includes the Note 15 provides clarification of what constitutes the Intake Structure Intake Cooling Waiter (ICW)       for the purpose of this EAL.
system and those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table R-3 Safe Operation & Shutdown Rooms/Areas Room/Area                                     Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)                     1, 2, 3, 4, 5 Turbine Building (east end levels required for access to 1, 2, 3, 4, 5 AFW, Steam Trestle, Switchgear and Control Rooms)
Cable Spreading Room                                                 1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                               1, 2, 3, 4, 5 Steam Trestle Area                                                   1, 2, 3, 4, 5 Intake Structure (Note 15)                                           1, 2, 3, 4, 5 Component Cooling Water Area                                         1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                                 1, 2, 3, 4, 5 Refueling Water Tank                                                 1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                                    1, 2, 3, 4 Containment                                                               4, 5 27 of 124

EAL Comparison Matrix PSL NEI IC#          NEI IC Wording                               PSL IC Wording                            Difference/Deviation Justification IC#(s)
AS1   Release of gaseous radioactivity   RS1   Release of gaseous radioactivity       None resulting in offsite dose greater        resulting in offsite dose greater than than 100 mrem TEDE or 500                100 mrem TEDE or 500 mrem thyroid mrem thyroid CDE                         CDE MODE: All                                MODE: All NEI Ex.                                    PSL NEI Example EAL Wording                          PSL EAL Wording                            Difference/Deviation Justification EAL #                                    EAL #
1    Reading on ANY of the following   RS1.1  Reading on any Table 1[2]R-1 effluent  The PSL radiation monitors that detect radioactivity effluent radiation monitors greater than          radiation monitor > column "SAE" for   release to the environment are listed in Table 1[2]R-1. UE, the reading shown for 15                   15 min. (Notes 1, 2, 3, 4)          Alert, SAE and GE thresholds for all PSL continuously minutes or longer:                                                              monitored gaseous and liquid release pathways are listed in (site-specific monitor list and                                                 Table 1[2]R-1 to consolidate the information in a single threshold values)                                                               location and, thereby, simplify identification of the thresholds by the EAL-user.
2    Dose assessment using actual      RS1.2  Dose assessment using actual           The site boundary is the site-specific receptor point.
meteorology indicates doses              meteorology indicates doses > 100 greater than 100 mrem TEDE or            mrem TEDE or 500 mrem thyroid CDE 500 mrem thyroid CDE at or               at or beyond the SITE BOUNDARY beyond (site-specific dose               (Note 4) receptor point) 3    Field survey results indicate     RS1.3  Field survey results indicate EITHER The site boundary is the site-specific receptor point.
EITHER of the following at or            of the following at or beyond the SITE beyond (site-specific dose               BOUNDARY:
receptor point):                            Closed window dose rates > 100 Closed window dose rates               mR/hr expected to continue for greater than 100 mR/hr                 60 min.
expected to continue for 60           Analyses of field survey samples minutes or longer.                     indicate thyroid CDE > 500 mrem Analyses of field survey               for 60 min. of inhalation.
samples indicate thyroid CDE greater than 500               (Notes 1, 2) 28 of 124

EAL Comparison Matrix mrem for one hour of inhalation.
Notes    The Emergency Director           Note 1: The EC should declare the     The classification timeliness note has been standardized should declare the Site Area           event promptly upon           across the PSL EAL scheme by referencing the "time limit" Emergency promptly upon                determining that the time     specified within the EAL wording.
determining that the                   limit has been exceeded, or applicable time has been               will likely be exceeded.
exceeded, or will likely be Note 2: If an ongoing release is exceeded.                                                            The classification timeliness note has been standardized detected and the release If an ongoing release is                                               across the PSL EAL scheme by referencing the "time limit" start time is unknown, detected and the release start                                       specified within the EAL wording.
assume that the release time is unknown, assume that duration has exceeded the the release duration has specified time limit.
exceeded 15 minutes.
If the effluent flow past an     Note 3: If the effluent flow past an effluent monitor is known to           effluent monitor is known to have stopped due to actions            have stopped due to actions   None to isolate the release path,           to isolate the release path, then the effluent monitor              then the effluent monitor reading is no longer valid for         reading is no longer VALID classification purposes.               for classification purposes.
The pre-calculated effluent monitor values presented in    Note 4  The pre-calculated effluent EAL #1 should be used for              monitor values presented in   Incorporated site-specific EAL numbers associated with EALs RA1.1, RS1.1 and         generic EAL#1.
emergency classification RG1.1 should be used for assessments until the results emergency classification from a dose assessment assessments until the results using actual meteorology are available.                             from a dose assessment using actual meteorology are available.
29 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                               PSL IC Wording                                 Difference/Deviation Justification IC#(s)
AS2    Spent fuel pool level at (site-    RS2   Spent fuel pool level at the top of the      Top of the fuel racks is the site-specific Level 3 description.
specific Level 3 description)            fuel racks MODE: All                                MODE: All NEI Ex.                                    PSL NEI Example EAL Wording                           PSL EAL Wording                                 Difference/Deviation Justification EAL #                                     EAL #
1     Lowering of spent fuel pool level RS2.1  Lowering of spent fuel pool level to 37      For PSL, Level 3 corresponds to 1 ft. above the top of the fuel to (site-specific Level 3 value)         ft. 3 in. [37 ft. 5 in.] (Level 3) on LI racks in the SFP, and is a reading of 37 ft. 3 in. [37 ft. 5 in.] on 1[2]A or LI-04-1[2]B                         LI-04-1[2]A or LI-04-1[2]B 30 of 124

EAL Comparison Matrix PSL NEI IC#           NEI IC Wording                              PSL IC Wording                              Difference/Deviation Justification IC#(s)
AG1    Release of gaseous radioactivity    RG1    Release of gaseous radioactivity   None resulting in offsite dose greater          resulting in offsite dose greater than 1,000 mrem TEDE or                    than 1,000 mrem TEDE or 5,000 5,000 mrem thyroid CDE.                    mrem thyroid CDE MODE: All                                  MODE: All NEI Ex.                                    PSL NEI Example EAL Wording                         PSL EAL Wording                             Difference/Deviation Justification EAL #                                     EAL #
1     Reading on ANY of the             RG1.1   Reading on any Table R-1 effluent  The PSL radiation monitors that detect radioactivity effluent following radiation monitors              radiation monitor > column "GE"    release to the environment are listed in Tables 1[2]R-1.
greater than the reading shown            for  15 min.                      UE, Alert, SAE and GE thresholds for all PSL continuously for 15 minutes or longer:                 (Notes 1, 2, 3, 4)                 monitored gaseous and liquid release pathways are listed in Table (site-specific monitor list and                                              1[2]R-1 to consolidate the information in a single location and, threshold values)                                                            thereby, simplify identification of the thresholds by the EAL-user.
2     Dose assessment using actual      RG1.2   Dose assessment using actual       The site boundary is the site-specific receptor point.
meteorology indicates doses               meteorology indicates doses >
greater than 1,000 mrem TEDE              1,000 mrem TEDE or 5,000 mrem or 5,000 mrem thyroid CDE at              thyroid CDE at or beyond the or beyond (site-specific dose             SITE BOUNDARY (Note 4) receptor point).
3     Field survey results indicate     RG1.3   Field survey results indicate      The site boundary is the site-specific field survey receptor point.
EITHER of the following at or             EITHER of the following at or beyond (site-specific dose                beyond the SITE BOUNDARY:
receptor point):
Closed window dose rates >
Closed window dose rates                    1,000 mR/hr expected to greater than 1,000 mR/hr                    continue for  60 min.
expected to continue for 60 minutes or longer.                        Analyses of field survey Analyses of field survey                    samples indicate thyroid CDE samples indicate thyroid                    > 5,000 mrem for 60 min. of CDE greater than 5,000                      inhalation.
31 of 124

EAL Comparison Matrix mrem for one hour of            (Notes 1, 2) inhalation.
Notes    The Emergency Director           Note 1:   The EC should declare     The classification timeliness note has been standardized across should declare the Site Area             the event promptly upon   the PSL EAL scheme by referencing the "time limit" specified Emergency promptly upon                  determining that the      within the EAL wording.
determining that the                     time limit has been applicable time has been                 exceeded, or will likely exceeded, or will likely be              be exceeded.
exceeded.
Note 2:   If an ongoing release is If an ongoing release is                   detected and the detected and the release                 release start time is     The classification timeliness note has been standardized across start time is unknown,                   unknown, assume that      the PSL EAL scheme by referencing the "time limit" specified assume that the release                  the release duration has  within the EAL wording.
duration has exceeded 15                 exceeded the specified minutes.                                 time limit.
If the effluent flow past an     Note 3:   If the effluent flow past effluent monitor is known to             an effluent monitor is have stopped due to actions              known to have stopped     None to isolate the release path,             due to actions to isolate then the effluent monitor                the release path, then reading is no longer valid for           the effluent monitor classification purposes.                 reading is no longer VALID for classification The pre-calculated effluent purposes for monitor values presented in classification purposes.
EAL #1 should be used for emergency classification       Note 4    The pre-calculated assessments until the results            effluent monitor values   Incorporated site-specific EAL numbers associated with generic from a dose assessment                   presented in EALs         EAL#1.
using actual meteorology are             RA1.1, RS1.1 and available.                               RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
32 of 124

EAL Comparison Matrix 33 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                               PSL IC Wording                                 Difference/Deviation Justification IC#(s)
AG2    Spent fuel pool level cannot be      RG2   Spent fuel pool level cannot be             Top of the fuel racks is the site-specific Level 3 description.
restored to at least (site-specific        restored to at least the top of the fuel Level 3 description) for 60                racks for 60 minutes or longer minutes or longer                          MODE: All MODE: All NEI Ex.                                      PSL NEI Example EAL Wording                             PSL EAL Wording                                 Difference/Deviation Justification EAL #                                       EAL #
1     Spent fuel pool level cannot be     RG2.1  Spent fuel pool level cannot be             For PSL, Level 3 corresponds to 1 ft. above the top of the fuel restored to at least (site-specific        restored to at least 37 ft. 3 in. [37 ft. 5 racks in the SFP, and is a reading of 37 ft. 3 in. [37 ft. 5 in.] on Level 3 value) for 60 minutes or           in.] (Level 3) on LI-04-1[2]A or LI     LI-04-1[2]A or LI-04-1[2]B longer                                     1[2]B for  60 min. (Note 1)
Note   The Emergency Director should        N/A   Note 1: The EC should declare the           The classification timeliness note has been standardized declare the General Emergency                       event promptly upon                across the PSL EAL scheme by referencing the "time limit" promptly upon determining that                      determining that the time limit    specified within the EAL wording.
60 minutes has been exceeded,                       has been exceeded, or will or will likely be exceeded.                         likely be exceeded.
34 of 124

EAL Comparison Matrix Category C Cold Shutdown / Refueling System Malfunction 35 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                       PSL IC Wording                                 Difference/Deviation Justification IC#(s)
CU1    UNPLANNED loss of (reactor       CU1   UNPLANNED loss of RCS                Deleted the words for 15 minutes or longer as the 15 minute vessel/RCS [PWR] or RPV                inventory                            criteria only applies to EAL #1
[BWR]) inventory for 15 minutes        MODE: 5 - Cold Shutdown, 6 -
or longer.                             Refueling MODE: Cold Shutdown, Refueling NEI Ex.                                  PSL NEI Example EAL Wording                    PSL EAL Wording                                  Difference/Deviation Justification EAL #                                  EAL #
1     UNPLANNED loss of reactor      CU1.1  UNPLANNED loss of reactor           None coolant results in (reactor           coolant results in RCS water vessel/RCS [PWR] or RPV               level < a required lower limit for
[BWR]) level less than a               15 min. (Note 1) required lower limit for 15 minutes or longer.
2                                    CU1.2  RCS water level cannot be           Added the words due to loss of RCS inventory to be consistent
: a. (Reactor vessel/RCS [PWR]
monitored                           with the IC wording.
or RPV [BWR]) level cannot AND EITHER be monitored.                                                          The Table C-1 sumps & tanks are the site-specific applicable sumps
* UNPLANNED increase in and tanks.
AND                                       any Table C-1 sump or tank level due to a loss of RCS      Although Visual observation is neither a sump nor tank, it is
: b. UNPLANNED increase in                  inventory                       included in order to implement the intent of the NEI basis which (site-specific sump and/or
* Visual observation of           states: operators may determine that an inventory loss is occurring tank) levels.                          UNISOLABLE RCS leakage          by observing changes Note    The Emergency Director should   N/A   Note 1: The EC should declare       The classification timeliness note has been standardized across the declare the Unusual Event                        the event promptly       PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                   upon determining that    EAL wording.
15 minutes has been exceeded,                    the time limit has been or will likely be exceeded.                      exceeded, or will likely be exceeded.
36 of 124

EAL Comparison Matrix Table C-1     Sumps/Tanks
* Containment Sump
* Reactor Cavity Sump
* Safeguards Sump
* Holdup Tanks
* Refueling Water Tank
* CCW Surge Tank
* Quench Tank
* Equipment Drain Tank 37 of 124

EAL Comparison Matrix PSL NEI IC#              NEI IC Wording                        PSL IC Wording                                  Difference/Deviation Justification IC#(s)
CU2    Loss of all but one AC power       CU2   Loss of all but one AC power          None source to emergency buses for            source to emergency buses for 15 minutes or longer.                    15 minutes or longer.
MODE: Cold Shutdown,                     MODE: 5 - Cold Shutdown, 6 -
Refueling, Defueled                      Refueling, D - Defueled NEI Ex.                                    PSL NEI Example EAL Wording                     PSL EAL Wording                                   Difference/Deviation Justification EAL #                                     EAL #
1                                       CU2.1  AC power capability, Table            4.16 kV emergency buses A3 and B3 are the PSL-specific
: a. AC power capability to (site-1[2]C-3, to 4.16 kV emergency         emergency buses.
specific emergency buses) is buses 1[2]A3 and 1[2]B3 reduced to a single power                                                  Table C-3 provides a consolidated list of AC power sources credited reduced to a single power source for 15 minutes or                                                   for this EAL.
source for  15 min. (Note 1) longer.
AND AND                                     Any additional single power source failure will result in loss of
: b. Any additional single power           all AC power to SAFETY source failure will result in        SYSTEMS loss of all AC power to SAFETY SYSTEMS.
Note    The Emergency Director should     N/A   Note 1: The EC should declare         The classification timeliness note has been standardized across the declare the Unusual Event                        the event promptly           PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                   upon determining that        EAL wording.
15 minutes has been exceeded,                    the time limit has been or will likely be exceeded.                      exceeded, or will likely be exceeded.
38 of 124

EAL Comparison Matrix Table 1[2]C-3 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer
* 1A[2A] Auxiliary Transformer back fed via Main Transformer (if already aligned)
* 1B[2B] Auxiliary Transformer back fed via Main Transformer (if already aligned)
Onsite
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned) 39 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                          PSL IC Wording                             Difference/Deviation Justification IC#(s)
CU3     UNPLANNED increase in RCS          CU3    UNPLANNED increase in RCS         None temperature                               temperature MODE: Cold Shutdown,                      MODE: 5 - Cold Shutdown, 6 -
Refueling                                 Refueling NEI Ex.                                     PSL NEI Example EAL Wording                       PSL EAL Wording                               Difference/Deviation Justification EAL #                                      EAL #
1     UNPLANNED increase in RCS          CU3.1  UNPLANNED increase in RCS          200&deg;F is the site-specific Tech. Spec. cold shutdown temperature temperature to greater than (site-        temperature to > 200&deg;F due to      limit.
specific Technical Specification          loss of decay heat removal         Added the words due to loss of decay heat removal capability cold shutdown temperature limit)          capability                         consistent with the generic bases.
2     Loss of ALL RCS temperature        CU3.2  Loss of all RCS temperature and    None and (reactor vessel/RCS [PWR]             RCS water level indication for or RPV [BWR]) level indication            15 min. (Note 1) for 15 minutes or longer.
Note   The Emergency Director should       N/A   Note 1: The EC should declare      The classification timeliness note has been standardized across the declare the Unusual Event                          the event promptly upon   PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                     determining that the time EAL wording.
15 minutes has been exceeded,                      limit has been exceeded, or will likely be exceeded                         or will likely be exceeded.
40 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                          PSL IC Wording                               Difference/Deviation Justification IC#(s)
CU4    Loss of Vital DC power for 15      CU4   Loss of vital DC power for 15        None minutes or longer.                       minutes or longer.
MODE: Cold Shutdown,                     MODE 5 - Cold Shutdown, 6 -
Refueling                                Refueling NEI Ex.                                     PSL NEI Example EAL Wording                        PSL EAL Wording                                Difference/Deviation Justification EAL #                                      EAL #
1     Indicated voltage is less than     CU4.1  Indicated voltage < 112 VDC on       The specified bus voltage indications are the minimum voltage (site-specific bus voltage value)         required vital 125 VDC buses         requirements for operability of the 125 VDC buses.
on required Vital DC buses for 15         1[2]A and 1[2]B for  15 min.        Vital 125 VDC battery buses 1[2]A and 1[2]B are the vital DC buses minutes or longer.                        (Note 1)                             credited for the EAL.
Note   The Emergency Director should       N/A   Note 1: The EC should declare        The classification timeliness note has been standardized across the declare the Unusual Event                         the event promptly upon      PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                    determining that the time    EAL wording.
15 minutes has been exceeded,                     limit has been exceeded, or will likely be exceeded.                       or will likely be exceeded.
41 of 124

EAL Comparison Matrix PSL NEI IC#           NEI IC Wording                         PSL IC Wording                            Difference/Deviation Justification IC#(s)
CU5    Loss of all onsite or offsite    CU5   Loss of all onsite or offsite   None communications capabilities.           communications capabilities.
MODE: Cold Shutdown,                   MODE: 5 - Cold Shutdown, 6 -
Refueling, Defueled                    Refueling, D - Defueled NEI Ex.                                   PSL NEI Example EAL Wording                     PSL EAL Wording                          Difference/Deviation Justification EAL #                                   EAL #
1    Loss of ALL of the following            Loss of all Table C-5 onsite CU5.1                                  Example EALs #1, 2 and 3 have been combined into a single onsite communication methods:           communication methods EAL for simplification of presentation.
OR (site specific list of                  Loss of all Table C-5 State and Table C-5 provides a site-specific list of onsite, offsite (ORO) and communications methods)                 local agency communication      NRC communications methods.
methods 2    Loss of ALL of the following ORO            OR communications methods:                 Loss of all Table C-5 NRC (site specific list of                  communication methods communications methods) 3    Loss of ALL of the following NRC communications methods:
(site specific list of communications methods) 42 of 124

EAL Comparison Matrix Table C-5 Communication Methods State/
System                      Onsite        NRC Local Plant Radio System                                   X Commercial Telephone System                          X      X     X Plant Address (PA) System                            X Federal Telephone System (ENS)                                    X EMnet                                                       X Hot Ring Down (HRD) Telephone System                        X Shift Manager Cell Phones                                   X     X 43 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                           PSL IC Wording                               Difference/Deviation Justification IC#(s)
CA1    Loss of (reactor vessel/RCS        CA1   Significant loss of RCS            Added the word Significant to differentiate the Alert loss of RCS
[PWR] or RPV [BWR]) inventory            inventory                          inventory IC from the Unusual Event IC which is Unplanned loss of MODE: Cold Shutdown,                     MODE: 5 - Cold Shutdown, 6 -       RCS inventory.
Refueling                                Refueling NEI Ex.                                    PSL NEI Example EAL Wording                       PSL EAL Wording                                Difference/Deviation Justification EAL #                                     EAL #
1     Loss of (reactor vessel/RCS       CA1.1   Loss of RCS inventory as           The minimum RCS level for continued LPSI (decay heat removal)
[PWR] or RPV [BWR]) inventory             indicated by RCS water level       operation is per 1[2]-AOP-03.02 Shutdown Cooling Abnormal as indicated by level less than           < minimum required for             Operations:
(site-specific level).                    continued LPSI pump operation
* Tygon hose 29 ft. 9 1/2 in.
* LI-1103 Level Cold Condition 0%
* LI-1117 RX REFUELING WATER LEVEL (NR) 24 1/2 in.
* LI-1117-1 RX REFUELING WATER LEVEL (WR) 24 1/2 in.
2     a. (Reactor vessel/RCS [PWR]      CA1.2   RCS water level cannot be          The Table C-1 sumps/tanks are the site-specific applicable sumps or RPV [BWR]) level cannot             monitored for  15 min. (Note 1)   and tanks.
be monitored for 15 minutes              AND EITHER                       Although Visual observation is neither a sump nor tank, it is or longer
* UNPLANNED increase in          included in order to implement the intent of the NEI basis which any Table C-1 sump or tank     states: operators may determine that an inventory loss is AND                                        level due to a loss of RCS     occurring by observing changes
: b. UNPLANNED increase in                     inventory (site-specific sump and/or
* Visual observation of tank) levels due to a loss of             UNISOLABLE RCS leakage (reactor vessel/RCS [PWR]
or RPV [BWR]) inventory.
Note   The Emergency Director should       N/A    Note 1: The EC should declare     The classification timeliness note has been standardized across the declare the Alert promptly upon                     the event promptly upon  PSL EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has                     determining that the     EAL wording.
been exceeded, or will likely be                    time limit has been 44 of 124

EAL Comparison Matrix exceeded     exceeded, or will likely be exceeded.
45 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                         PSL IC Wording                                 Difference/Deviation Justification IC#(s)
CA2    Loss of all offsite and all onsite  CA2   Loss of all offsite and all onsite None AC power to emergency buses               AC power to emergency buses for 15 minutes or longer                  for 15 minutes or longer.
MODE: Cold Shutdown,                      MODE: 5 - Cold Shutdown, 6 -
Refueling, Defueled                       Refueling, DEF - Defueled NEI Ex.                                     PSL NEI Example EAL Wording                       PSL EAL Wording                                 Difference/Deviation Justification EAL #                                      EAL #
1     Loss of ALL offsite and ALL        CA2.1  Loss of all offsite and all onsite 4.16 kV emergency buses 1[2]A3 and 1[2]B3 are the PSL-specific onsite AC Power to (site-specific         AC power to 4.16 kV emergency      emergency buses.
emergency buses) for 15                   buses 1[2]A3 and 1[2]B3 for minutes or longer.                        15 min. (Note 1)
Note   The Emergency Director should       N/A   Note 1: The EC should declare      The classification timeliness note has been standardized across the declare the Unusual Event                          the event promptly        PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                     upon determining that     EAL wording.
15 minutes has been exceeded,                      the time limit has been or will likely be exceeded.                        exceeded, or will likely be exceeded.
46 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                           PSL IC Wording                                 Difference/Deviation Justification IC#(s)
CA3    Inability to maintain the plant in  CA3   Inability to maintain plant in cold  None cold shutdown.                            shutdown.
MODE: Cold Shutdown,                      MODE: 5 - Cold Shutdown, 6 -
Refueling                                 Refueling NEI Ex.                                      PSL NEI Example EAL Wording                        PSL EAL Wording                                Difference/Deviation Justification EAL #                                      EAL #
1     UNPLANNED increase in RCS                 UNPLANNED increase in RCS           Example EALs #1 and #2 have been combined into a single EAL temperature to greater than               temperature to > 200&deg;F for          as EAL #2 is the alternative threshold based on a loss of RCS (site-specific Technical                  > Table C-4 duration due to a       temperature indication.
Specification cold shutdown               loss of RCS cooling (Notes 1, 13)                       Added due to a loss of RCS cooling consistent with the generic temperature limit) for greater                                                bases that states this EAL addresses conditions involving a loss of than the duration specified in                OR UNPLANNED RCS pressure             decay heat removal capability.
the following table.
increase > 10 psi (does not        200&deg;F is the site-specific Tech. Spec. cold shutdown temperature CA3.1   apply to solid plant conditions) 2     UNPLANNED RCS pressure                                                        limit.
increase greater than (site-Table C-4 is the site-specific implementation of the generic RCS specific pressure reading).
Reheat Duration Threshold table.
(This EAL does not apply during water-solid plant conditions.                                                 10 psig is the site-specific RCS pressure increase readable by
[PWR])                                                                        Control Room indications.
Note    The Emergency Director should       N/A   Note 1: The EC should declare       The classification timeliness note has been standardized across the declare the Unusual Event                            the event promptly       PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                       upon determining that    EAL wording.
15 minutes has been exceeded,                        the time limit has been or will likely be exceeded.                          exceeded, or will likely be exceeded.
47 of 124

EAL Comparison Matrix N/A    N/A                               N/A       Note 13: If an RCS heat                  Added Note 13 consistent with the asterisk note provided in the removal system is in           generic RCS Heat-up Duration Threshold table.
operation within the applicable Table C-4 heat-up duration and RCS temperature is being reduced, the EAL is not applicable.
Table: RCS Heat-up Duration Thresholds RCS Status                      Containment Closure Status                 Heat-up Duration Intact (but not at reduced inventory Not applicable                          60 minutes*
[PWR])
Not intact (or at reduced inventory                   Established                           20 minutes*
[PWR])                                             Not Established                           0 minutes
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
Table C-4     RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Status                                                                Heat-up Duration Status Intact AND not reduced N/A                                 60 min.
inventory Established                               20 min.
Not intact OR reduced inventory Not established                              0 min.
48 of 124

EAL Comparison Matrix PSL NEI IC#          NEI IC Wording                     PSL IC Wording                          Difference/Deviation Justification IC#(s)
CA6    Hazardous event affecting a  CA6   Hazardous event affecting     Revised wording from affecting a SAFETY SYSTEM to read SAFETY SYSTEM needed for           SAFETY SYSTEMS needed for     affecting SAFETY SYSTEMS to align with changes made the current operating mode.        the current operating mode. consistent with NRC EP FAQ 2016-002.
MODE: Cold Shutdown,               MODE: 5 - Cold Shutdown, 6 -
Refueling                          Refueling 49 of 124

EAL Comparison Matrix NEI Ex.                                   PSL NEI Example EAL Wording                    PSL EAL Wording                            Difference/Deviation Justification EAL #                                   EAL #
1    a. The occurrence of ANY of      CA6.1                               The hazardous events have been tabularized in Table C-6.
The occurrence of any Table the following hazardous C-6 hazardous event          The proposed PSL CA6.1 and SA9.1 wording is intended to ensure events:
Seismic event                        AND                       that an Alert should be declared only when actual or potential (earthquake)                                                  performance issues with SAFETY SYSTEMS have occurred as a Event damage has caused       result of a hazardous event. The occurrence of a hazardous event Internal or external indications of degraded       will result in an Unusual Event classification at a minimum. In order flooding event performance on one train of a to warrant escalation to the Alert classification, the hazardous event High winds or tornado strike                          SAFETY SYSTEM needed for      should cause indications of degraded performance to one train of a FIRE the current operating mode    SAFETY SYSTEM with either indications of degraded performance EXPLOSION                            AND EITHER:               on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the (site-specific hazards)                                        second SAFETY SYSTEM train, such that the operability or Event damage has        reliability of the second train is a concern. In addition, escalation to Other events with similar hazard characteristics as             caused indications of   the Alert classification should not occur if the damage from the determined by the Shift               degraded performance to hazardous event is limited to a SAFETY SYSTEM that was Manager                               the second train of the inoperable, or out of service, prior to the event occurring. As such, AND                                SAFETY SYSTEM           the proposed EALs will reduce the potential of declaring an Alert
: b. EITHER of the following:                 needed for the current  when events are in progress that do not involve an actual or
: 1. Event damage has caused               operating mode          potential substantial degradation of the level of safety of the plant, indications of degraded              Event damage has i.e., does not cause significant concern with shutting down or performance in at least               resulted in VISIBLE     cooling down the plant.
one train of a SAFETY                 DAMAGE to the second    EALs CA6.1 and SA9.1 do not directly escalate to a Site Area SYSTEM needed for the                 train of the SAFETY     Emergency or a General Emergency due to a hazardous event.
current operating mode.               SYSTEM needed for the   The Fission Product Barrier and/or Abnormal Radiation OR                                 current operating mode  Levels/Radiological Effluent recognition categories would provide
: 2. The event has caused                                          an escalation path to a Site Area Emergency or a General VISIBLE DAMAGE to a             (Notes 9, 10)
Emergency.
SAFETY SYSTEM component or structure                                      The EALs and the Basis sections have been revised to ensure needed for the current                                      potential escalations from an Unusual Event to an Alert, due to a operating mode.                                             hazardous event, is appropriate as the concern with these EALs is:
(1) a hazardous event has occurred, (2) one SAFETY SYSTEM train is having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or the VISIBLE DAMAGE is enough to be concerned that the second SAFETY SYSTEM train may have 50 of 124

EAL Comparison Matrix operability or reliability issues.
The definition for VISIBLE DAMAGE has been revised to reflect the fact that the EALs are based upon SAFETY SYSTEM trains rather than individual components or structures.
Note 9 has been added to CA6.1 and SA9.1 as it meets the intent of the EALs, is consistent with other EALs (e.g., EAL HA5.1 which was previously endorsed by the NRC), and ensures that declared emergencies are based upon unplanned events with the potential to pose a radiological risk to the public.
Note 10 has been added to CA6.1 and SA9.1 to help reinforce and succinctly capture the more detailed information from the revised basis section related to when conditions would require the declaration of an Alert.
CA6.1 and SA9.1 are consistent with NRC FAQ 2016-002 requiring degraded performance or visible damage to more than one safety system train caused by the specified events.
This revised wording is a deviation from the NEI 99-01 Revision 6 CA6 and SA9 generic wording and bases but is deemed acceptable consistent with endorsed NRC EP FAQ 2016-002.
51 of 124

EAL Comparison Matrix N/A   N/A           N/A Note 9: If the affected SAFETY    Added Note 9 consistent with the recommendation of NRC EP FAQ SYSTEM train was         2016-002.
already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.
Note 10: If the hazardous event   Added Note 10 consistent with the recommendation of NRC EP only resulted in         FAQ 2016-002.
VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
52 of 124

EAL Comparison Matrix Table C-6       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 53 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                         PSL IC Wording                                Difference/Deviation Justification IC#(s)
CS1    Loss of (reactor vessel/RCS       CS1   Loss of RCS inventory affecting     None
[PWR] or RPV [BWR]) inventory           core decay heat removal affecting core decay heat               capability removal capability.                     MODE: 5 - Cold Shutdown, 6 -
MODE: Cold Shutdown,                    Refueling Refueling NEI Ex.                                    PSL NEI Example EAL Wording                      PSL EAL Wording                               Difference/Deviation Justification EAL #                                    EAL #
1    a. CONTAINMENT CLOSURE                   0 in. on LI-1117 or LI-1117-1      0 in. on LI-1117 or LI-1117-1 is the lowest RCS level that can be not established.                                                        monitored in the cold condition and corresponds to the bottom of the RCS hot leg. This level is therefore used, in lieu of 6 in. below AND CS1.1                                     the bottom of the hot leg as specified in NEI 99-01 Rev. 6, for this
: b. (Reactor vessel/RCS [PWR]                                                example EAL at PSL.
or RPV [BWR]) level less than (site-specific level).
2    a. CONTAINMENT CLOSURE                                                      0 in. on LI-1117 or LI-1117-1 is the lowest RCS level that can be established.                                                            monitored in the cold condition and corresponds to the bottom of the RCS hot leg. Therefore this example EAL based on RCS level AND                                                                     below TAF, consistent with the developers notes, is not N/A                   N/A
: b. (Reactor vessel/RCS [PWR]                                               implemented for PSL.
or RPV [BWR]) level less than (site-specific level).
3     a. (Reactor vessel/RCS [PWR]     CS1.2  RCS level cannot be monitored      Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot           for  30 min. (Note 1)             improve the readability of the EAL.
be monitored for 30 minutes               AND                           Although Visual observation is neither a sump nor tank, it is or longer.                           Core uncovery is indicated by      included in order to implement the intent of the NEI basis which any of the following:
AND                                                                     states: operators may determine that an inventory loss is
* UNPLANNED increase in           occurring by observing changes
: b. Core uncovery is indicated by              any Table C-1 sump or tank ANY of the following:                     level of sufficient magnitude In the Refueling mode, as water level in the reactor vessel lowers, 54 of 124

EAL Comparison Matrix (Site-specific radiation           to indicate core uncovery     the dose rate above the core will rise. The dose rate due to this monitor) reading greater
* Visual observation of          core shine should result in Containment High Range Radiation than (site-specific value)         UNISOLABLE RCS leakage        Monitor readings of > 4.8E+04 R/hr when RCS water level drops to of sufficient magnitude to    the top of the core.
Erratic source range indicate core uncovery        No other site-specific indications of core uncovery have been monitor indication [PWR]
* Containment High Range         identified for PSL.
UNPLANNED increase in              Radiation Monitor reading (site-specific sump and/or         > 4.8E+04 R/hr (Refueling tank) levels of sufficient         Mode) magnitude to indicate
* Erratic source range core uncovery                      monitor indications (Other site-specific indications)
Note    The Emergency Director should     N/A Note 1: The EC should declare     The classification timeliness note has been standardized across the declare the Site Area                          the event promptly upon  PSL EAL scheme by referencing the "time limit" specified within the Emergency promptly upon                        determining that the     EAL wording.
determining that 30 minutes has                time limit has been been exceeded, or will likely be               exceeded, or will likely exceeded                                       be exceeded.
55 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                           PSL IC Wording                              Difference/Deviation Justification IC#(s)
CG1    Loss of (reactor vessel/RCS         CG1    Loss of RCS inventory affecting   None
[PWR] or RPV [BWR]) inventory              fuel clad integrity with affecting fuel clad integrity with         containment challenged containment challenged                     MODE: 5 - Cold Shutdown, 6 -
MODE: Cold Shutdown,                       Refueling Refueling NEI Ex.                                      PSL NEI Example EAL Wording                        PSL EAL Wording                           Difference/Deviation Justification EAL #                                      EAL #
1    a. (Reactor vessel/RCS [PWR]                                                 0 in. on LI-1117 or LI-1117-1 is the lowest RCS level that can be or RPV [BWR]) level less than                                            monitored in the cold condition and corresponds to the bottom of (site-specific level) for 30                                             the RCS hot leg. Therefore this example EAL based on RCS minutes or longer.                                                       level below TAF, consistent with the developers notes, is not N/A                   N/A AND                                                                      implemented for PSL.
: b. ANY indication from the Containment Challenge Table (see below).
2    a. (Reactor vessel/RCS [PWR]        CG1.1  RCS level cannot be monitored     Site-specific applicable sumps and tanks are listed in Table C-1 or RPV [BWR]) level cannot             for  30 min. (Note 1)            to improve the readability of the EAL.
be monitored for 30 minutes              AND                             Although Visual observation is neither a sump nor tank, it is or longer.                             Core uncovery is indicated by     included in order to implement the intent of the NEI basis which any of the following:
AND                                                                    states: operators may determine that an inventory loss is
* UNPLANNED increase in          occurring by observing changes
: b. Core uncovery is indicated by               any Table C-1 sump or tank ANY of the following:                      level of sufficient magnitude In the Refueling mode, as water level in the reactor vessel to indicate core uncovery     lowers, the dose rate above the core will rise. The dose rate due (Site-specific radiation                                            to this core shine should result in Containment High Range
* Visual observation of monitor) reading greater                                            Radiation Monitor readings of > 4.8E+04 R/hr when RCS water UNISOLABLE RCS leakage than (site-specific value)                                          level drops to the top of the core.
of sufficient magnitude to Erratic source range                  indicate core uncovery        No other site-specific indications of core uncovery have been monitor indication [PWR]
* Containment High Range        identified for PSL.
Radiation Monitor reading >
UNPLANNED increase in                                               4% hydrogen concentration in the presence of oxygen is the (site-specific sump and/or             4.8E+04 R/hr (Refueling 56 of 124

EAL Comparison Matrix tank) levels of sufficient          Mode)                          minimum necessary to support a hydrogen explosion.
magnitude to indicate core
* Erratic source range monitor uncovery                           indications AND (Other site-specific Any Containment Challenge indications) indication, Table C-2 AND
: c. ANY indication from the Containment Challenge Table (see below).
Note   The Emergency Director should       N/A Note 1: The EC should declare       The classification timeliness note has been standardized across declare the General Emergency                    the event promptly upon    the PSL EAL scheme by referencing the "time limit" specified promptly upon determining that                   determining that the time  within the EAL wording.
30 minutes has been exceeded,                    limit has been exceeded, or will likely be exceeded.                      or will likely be exceeded.
Note 6: If CONTAINMENT              Note 6 implements the asterisked note associated with the CLOSURE is re-             Containment Closure requirement.
established prior to exceeding the 30-min.
time limit, declaration of a General Emergency is not required.
57 of 124

EAL Comparison Matrix Containment Challenge Table CONTAINMENT CLOSURE not established*
(Explosive mixture) exists inside containment UNPLANNED increase in containment pressure Secondary containment radiation monitor reading above (site-specific value) [BWR]
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
Table C-3 Containment Challenge Indications
* CONTAINMENT CLOSURE not established (Note 6)
* CTMT hydrogen concentration > 4%
* UNPLANNED increase in CTMT pressure 58 of 124

EAL Comparison Matrix Category D Permanently Defueled Station Malfunction 59 of 124

EAL Comparison Matrix PSL NEI IC#          NEI IC Wording                    PSL IC Wording                     Difference/Deviation Justification IC#(s)
PD-AU1       Recognition Category D      N/A   N/A                   NEI Recognition Category PD ICs and EALs are applicable only to PD-AU2     Permanently Defueled Station                              permanently defueled stations. PSL is not a defueled station.
PD-SU1 PD-HU1 PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 60 of 124

EAL Comparison Matrix Category E Independent Spent Fuel Storage Installation 61 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                           PSL IC Wording                            Difference/Deviation Justification IC#(s)
E-HU1   Damage to a loaded cask             EU1    Damage to a loaded cask           None CONFINEMENT BOUNDARY                       CONFINEMENT BOUNDARY MODE: All                                  MODE: All NEI Ex.                                    PSL NEI Example EAL Wording                       PSL EAL Wording                           Difference/Deviation Justification EAL #                                    EAL #
1    Damage to a loaded cask           EU1.1   Damage to a loaded cask          The specified EAL threshold values correspond to 2 times the CONFINEMENT BOUNDARY as                   CONFINEMENT BOUNDARY             Horizontal Storage Module (HSM-H) external surface dose rate indicated by an on-contact                as indicated by an on-contact    limits.
radiation reading greater than (2         radiation reading on the surface times the site-specific cask              of a loaded spent fuel cask specific technical specification          (HSM) > any of the following:
allowable radiation level) on the surface of the spent fuel cask.
* 1,600 mrem/hr on the HSM front bird screen
* 10 mrem/hr on the outside HSM door centerline
* 10 mrem/hr on the HSM end shield wall exterior 62 of 124

EAL Comparison Matrix Category F Fission Product Barrier Degradation 63 of 124

EAL Comparison Matrix PWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI                                       PSL NEI Threshold Wording                        PSL FPB Wording                                Difference Justification FPB#                                     FPB #(s)
FC Loss  RCS or SG Tube Leakage             N/A    N/A                             N/A 1     Not Applicable FC Loss  Inadequate Heat Removal          FC Loss  1. REP CETs > 1200&deg;F            Core exit thermocouple readings > 1200&deg;F is the site-specific temperature value for PSL.
2     A. Core exit thermocouple          B.1 readings greater than (site-specific temperature value).
FC Loss  RCS Activity/CMNT Rad            FC Loss  2. CTMT high range radiation    CTMT high range radiation monitors Ch. # 58/59 are the C.2       monitors Ch. # 58/59         Containment High Range area radiation monitors. The threshold 3     A. Containment radiation                     [RIM-26-40/41]               values specified has been calculated assuming the monitor reading greater than                 > 1.4E+02 R/hr               instantaneous release and dispersal of the reactor coolant noble (site-specific value) gas and iodine inventory associated with approximately 1% fuel OR                                                                     clad damage.
B. (Site-specific indications FC Loss  3. Coolant activity > 300       None that reactor coolant activity is C.3         &#xb5;Ci/gm dose equivalent greater than 300 &#xb5;Ci/gm dose I-131 equivalent I-131)
FC Loss  CNMT Integrity or Bypass           N/A    N/A                             N/A 4     Not Applicable FC Loss  Other Indications                  N/A    N/A                             No other site-specific Fuel Clad Loss indication has been identified 5                                                                               for PSL.
A. (site-specific as applicable) 64 of 124

EAL Comparison Matrix NEI                                          PSL NEI Threshold Wording                           PSL FPB Wording                                  Difference Justification FPB#                                        FPB #(s)
FC Loss  ED Judgment                         FC Loss  4. Any condition in the opinion  None E.4       of the EC that indicates loss 6    A. ANY condition in the of the Fuel Clad barrier opinion of the Emergency Director that indicates Loss of the Fuel Clad Barrier.
FC    RCS or SG Tube Leakage              FC Pot. RVLMS indicates Sensors 4        PSL includes in its Inadequate Core Cooling instrumentation a P-Loss                                       Loss    through 8 not covered (Note 12)  reactor vessel level monitoring system (RVLMS) that is displayed A. RCS/reactor vessel level A.1                                     to the operators and can measure discrete reactor vessel water 1        less than (site-specific level) levels from the top of the reactor vessel head to the top of the fuel alignment plate (Sensor #8). The bottom of this instrument's span in the reactor vessel plenum is the lowest available reactor vessel level indicator and is used in this EAL to represent approximately the top of active fuel.
FC    Inadequate Heat Removal             FC Pot. 2. REP CETs > 700&deg;F              Core exit thermocouple readings > 700&deg;F is the site-specific P-Loss                                       Loss                                     temperature value for PSL.
A. Core exit thermocouple B.2 2    readings greater than (site-specific temperature value)
OR FC Pot. 3. Both SG wide range level      Wide range level in both steam generators less than 15% for  15 B. Inadequate RCS heat               Loss        < 15% for  15 min.          minutes without Once-Through Core Cooling (OTC) established removal capability via steam          B.3                                     indicates an extreme challenge to the ability to remove RCS heat AND generators as indicated by                                                    using the steam generators (i.e., loss of an effective secondary-(site-specific indications).                     OTC not established          side heat sink). The 15-minute criteria provides the appropriate time for establishment of the heat removal safety function and aligns with the required Functional Recovery safety function status check frequency.
FC    RCS Activity/CMNT Rad                 N/A    N/A                              N/A P-Loss Not Applicable 3
65 of 124

EAL Comparison Matrix NEI                                       PSL NEI Threshold Wording                        PSL FPB Wording                                   Difference Justification FPB#                                     FPB #(s)
FC    CNMT Integrity or Bypass           N/A    N/A                                N/A P-Loss Not Applicable 4
FC    Other Indications                  N/A    N/A                                No other site-specific Fuel Clad Potential Loss indication has been P-Loss                                                                               identified for PSL.
A. (site-specific as applicable) 5 FC    Emergency Director               FC Pot. 4. Any condition in the opinion    None P-Loss  Judgment                          Loss       of the EC that indicates E.4       potential loss of the Fuel Clad 6    A. Any condition in the opinion barrier.
of the Emergency Director that indicates Potential Loss of the Fuel Clad Barrier.
66 of 124

EAL Comparison Matrix PWR RCS Fission Product Barrier Degradation Thresholds NEI                                     PSL FPB NEI IC Wording                          PSL FPB Wording                                Difference Justification FPB#                                       #(s)
RCS    RCS or SG Tube Leakage           RCS Loss    1. An automatic or manual     SIAS is the site-specific terminology for ECCS (SI) actuation.
Loss                                       A.1         SIAS actuation required by A. An automatic or manual 1         ECCS (SI) actuation is                     EITHER:
required by EITHER of the
* UNISOLABLE RCS following:                                    leakage
: 1. UNISOLABLE RCS
* SG tube RUPTURE leakage OR
: 2. SG tube RUPTURE.
RCS    Inadequate Heat Removal             N/A      N/A                           N/A Loss Not Applicable 2
RCS    RCS Activity/CMNT Rad            RCS Loss    2. Any CIAS [CIS] monitor Ch. The specified radiation monitor reading corresponds to an Loss                                       C.2         # 3/4/5/6 [RIM-26-3/4/5/6] instantaneous release of all reactor coolant mass into the A. Containment radiation 3                                                    > 1.5E+03 mR/hr            containment, assuming that reactor coolant activity equals monitor reading greater than Technical Specification allowable limits (0.1% failed fuel) on the (site-specific value).
specified radiation monitors.
RCS    CNMT Integrity or Bypass            N/A      N/A                           N/A Loss Not Applicable 4
RCS    Other Indications                   N/A      N/A                           No other site-specific RCS Loss indication has been identified for Loss                                                                              PSL.
A. (site-specific as applicable) 5 RCS    Emergency Director Judgment      RCS Loss    3. Any condition in the       None opinion of the EC that 67 of 124

EAL Comparison Matrix NEI                                           PSL FPB NEI IC Wording                              PSL FPB Wording                              Difference Justification FPB#                                             #(s)
Loss   A. ANY condition in the opinion           E.3      indicates loss of the RCS of the Emergency Director that                     barrier 6
indicates Loss of the RCS Barrier.
RCS    RCS or SG Tube Leakage                 RCS Pot. 1. UNISOLABLE RCS or SG      PSL has implemented the alternative RCS potential loss threshold Loss       tube leakage > 50 gpm     provided in the generic guidance developer notes. Starting of a P-Loss 1 A. Operation of a standby A.1                                standby charging pump is not representative of RCS leak size charging (makeup) pump is relative to charging pump capacity.
required by EITHER of the following:
: 1. UNISOLABLE RCS leakage OR                                                                 The maximum allowed RCS cooldown rate per Technical RCS Pot. 2. Uncontrolled RCS cooldown Specifications is 100&#xba;F/hr with cold leg temperature > 105&#xba;F.
: 2. SG tube leakage.               Loss      > 100&#xba;F/hr and RCS A.2      pressure and temperature  A RCS cooldown greater than 100&deg;F/hr below 500&deg;F is the OR                                           are to the left of the    temperature that requires implementation of Pressurized Thermal B. RCS cooldown rate greater                    Maximum Subcooled (PT)    Shock (PTS) guidance.
than (site-specific                          Curve (EOP Figure 1A or pressurized thermal shock                    1B criteria/limits defined by site-specific indications).
RCS    Inadequate Heat Removal                RCS Pot. 3. Both SG wide range level  Wide range level in both steam generators less than 15% for  15 Loss       < 15% for  15 min.       minutes without Once-Through Core Cooling (OTC) established P-Loss 2 A. Inadequate RCS heat B.3                                indicates an extreme challenge to the ability to remove RCS heat removal capability via steam                    AND using the steam generators (i.e., loss of an effective secondary-generators as indicated by OTC not established       side heat sink). The 15-minute criteria provides the appropriate (site-specific indications).
time for establishment of the heat removal safety function and aligns with the required Functional Recovery safety function status check frequency.
RCS    RCS Activity/CMNT Rad                    N/A    N/A                          N/A P-Loss 3 Not Applicable 68 of 124

EAL Comparison Matrix NEI                                     PSL FPB NEI IC Wording                          PSL FPB Wording                                 Difference Justification FPB#                                       #(s)
RCS    CNMT Integrity or Bypass           N/A    N/A                             N/A P-Loss 4 Not Applicable RCS    Other Indications                  N/A    N/A                             No other site-specific RCS Potential Loss indication has been identified for PSL.
P-Loss 5 A. (site-specific as applicable)
RCS    Emergency Director Judgment      RCS Pot. 4. Any condition in the opinion None Loss       of the EC that indicates P-Loss 6 A. ANY condition in the opinion E.4       potential loss of the RCS of the Emergency Director that barrier indicates Potential Loss of the RCS Barrier.
69 of 124

EAL Comparison Matrix PWR Containment Fission Product Barrier Degradation Thresholds NEI                                            PSL NEI IC Wording                                 PSL FPB Wording                              Difference Justification FPB#                                          FPB #(s)
CNMT   RCS or SG Tube Leakage                  CTMT    1. A leaking or RUPTURED SG is         None Loss                                          Loss       FAULTED outside of CTMT A. A leaking or RUPTURED SG is A.1 1    FAULTED outside of containment.
CNMT   Inadequate Heat Removal                  N/A    N/A                                    N/A Loss Not Applicable 2
CNMT   RCS Activity/CMNT Rad                    N/A    N/A                                    N/A Loss Not applicable 3
CNMT   CNMT Integrity or Bypass                CTMT    2. CTMT isolation (CIS [CIAS]) is      Specified (CIS [CIAS]) as the site-specific isolation Loss                                          Loss       required                            signals that initiate required CTMT isolations to A. Containment isolation is required D.2       AND EITHER:                         preclude fission product release from CTMT.
4 AND
* CTMT integrity has been lost   Added the word atmosphere to the second bulleted EITHER of the following:                           based on EC judgment           threshold to reinforce the generic bases that the
* UNISOLABLE pathway from        intent is an unisolable pathway from the containment
: 1. Containment integrity has                       CTMT atmosphere to the         atmosphere, not RCS. RCS leakage outside been lost based on                             environment exists             containment is addressed under CTMT Loss D.3 Emergency Director judgment.                                                  below.
OR CTMT    3. Indications of RCS leakage outside  None
: 2. UNISOLABLE pathway from Loss       of CTMT the containment to the D.3 environment exists.
OR B. Indications of RCS leakage outside of containment.
70 of 124

EAL Comparison Matrix NEI                                                PSL NEI IC Wording                                     PSL FPB Wording                                Difference Justification FPB#                                              FPB #(s)
CNMT   Other Indications                            N/A    N/A                                       No other site-specific containment Loss indication Loss                                                                                                has been identified for PSL.
A. (site-specific as applicable) 5 CNMT   Emergency Director Judgment                 CTMT    4. Any condition in the opinion of the    None Loss                                              Loss        EC that indicates loss of the CTMT ANY condition in the opinion of the 6                                                 E.4        barrier Emergency Director that indicates Loss of the Containment Barrier.
CNMT   RCS or SG Tube Leakage                       N/A    N/A                                       N/A P-Loss Not Applicable 1
CNMT   Inadequate Heat Removal                     CTMT    1. REP CETs > 1200&deg;F                    Core Damage Assessment uses a CET temperature P-Loss                                               Pot.                                             of 1200 &deg;F as the temperature indicative of a A. 1. (Site-specific criteria for entry                 AND Loss                                              significantly damaged core.
2              into core cooling restoration B.1        Functional Recovery (1[2]-EOP-15) procedure)                                                                          Functional Recovery (1[2]-EOP-15) is the site specific for RCS and Core Heat Removal restoration procedure for PSL.
AND                                           not effective within 15 min. (Note 1)
: 2. Restoration procedure not effective within 15 minutes.
CNMT   RCS Activity/CMNT Rad                       CTMT    2. CTMT high range radiation              CTMT high range radiation monitors Ch. # 58/59 P-Loss                                               Pot.      monitors Ch. # 58/59 [RIM          [RIM-26-40/41] are the containment High Range area A. Containment radiation monitor Loss       40/41] > 2.7E+03 R/hour                radiation monitors. The radiation monitor readings 3        reading greater than (site-specific C.2                                              specified correspond to an instantaneous release of value).
all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed.
CNMT   CNMT Integrity or Bypass                    CTMT    3. CTMT pressure > 42 psig                The containment design pressure is 42 psig.
P-Loss                                               Pot.
A. Containment pressure greater Loss 4         than (site-specific value)
D.3 71 of 124

EAL Comparison Matrix NEI                                            PSL NEI IC Wording                               PSL FPB Wording                               Difference Justification FPB#                                          FPB #(s)
OR                                                                                 A containment hydrogen concentration of 4%
CTMT    4. CTMT hydrogen concentration B. Explosive mixture exists inside       Pot.      > 4%                                conservatively represents the lowest threshold for containment                         Loss                                           flammability in the presence of oxygen.
D.4 OR C. 1. Containment pressure greater than (site-specific pressure setpoint)
CTMT    5. CTMT pressure > 10 psig [5.4 psig]  The containment pressure setpoint 10 psig [5.4 psig]
AND                            Pot.      with < one full train of CTMT heat  is the pressure at which the containment Loss       removal systems operating per       depressurization equipment should actuate and begin
: 2. Less than one full train of D.5       design for  15 min. (Notes 1, 11)  performing its function.
(site-specific system or equipment) is operating per                                                   Added Note 1 consistent with other thresholds with a design for 15 minutes or                                                      timing component.
longer.
Added Note 11 to define what constitutes a full train of containment heat removal systems.
CNMT   Other Indications                        N/A                      N/A                  No other site-specific containment Potential Loss P-Loss                                                                                         indication has been identified for PSL.
A. (site-specific as applicable) 5 CNMT   Emergency Director Judgment             CTMT    6. Any condition in the opinion of the None P-Loss                                           Pot.      EC that indicates potential loss of A. ANY condition in the opinion of the Loss       the CTMT barrier 6    Emergency Director that indicates E.6 Potential Loss of the Containment Barrier.
72 of 124

EAL Comparison Matrix Category H Hazards and Other Conditions Affecting Plant Safety 73 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                            PSL IC Wording                              Difference/Deviation Justification IC#(s)
HU1   Confirmed SECURITY                   HU1   Confirmed SECURITY                    None CONDITION or threat                        CONDITION or threat.
MODE: All                                  MODE: All NEI Ex.                                      PSL NEI Example EAL Wording                          PSL EAL Wording                             Difference/Deviation Justification EAL #                                      EAL #
1    A SECURITY CONDITION that           HU1.1   A SECURITY CONDITION that           Example EALs #1, 2 and 3 have been combined into a single EAL does not involve a HOSTILE                  does not involve a HOSTILE          for ease of presentation and use.
ACTION as reported by the (site-            ACTION as reported by PSL specific security shift                     Security Shift Supervision supervision).                                  OR Notification of a credible security 2    Notification of a credible security         threat directed at the site threat directed at the site.                   OR A validated notification from the 3    A validated notification from the           NRC providing information of an NRC providing information of an             aircraft threat aircraft threat.
74 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                          PSL IC Wording                             Difference/Deviation Justification IC#(s)
HU2    Seismic event greater than OBE     HU2    Seismic event greater than OBE   None levels                                    levels MODE: All                                 MODE: All NEI Ex.                                    PSL NEI Example EAL Wording                      PSL EAL Wording                           Difference/Deviation Justification EAL #                                    EAL #
1     Seismic event greater than       HU2.1   Seismic event > OBE as          0-AOP-117.02 Earthquake provides the guidance for determining Operating Basis Earthquake               indicated by S-36[S-44] CNTMT   if the OBE earthquake threshold is exceeded (horizontal or (OBE) as indicated by:                   SEISMIC MONITOR HIGH            vertical) and any required response actions. The Control Rooms LIMIT alarm                     have real time OBE exceedance alarm indications (S-36[S-44]
(site-specific indication that a CNTMT SEISMIC MONITOR HIGH LIMIT alarm).
seismic event met or exceeded OBE limits)                                                              For both Unit 1 and Unit 2, the OBE ground acceleration threshold is > 0.05g horizontal and vertical.
75 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                            PSL IC Wording                            Difference/Deviation Justification IC#(s)
HU3    Hazardous event.                     HU3   Hazardous event                   None MODE: All                                  MODE: All NEI Ex.                                      PSL NEI Example EAL Wording                         PSL EAL Wording                            Difference/Deviation Justification EAL #                                       EAL #
1     A tornado strike within the         HU3.1  A tornado strike within the       None PROTECTED AREA.                            PROTECTED AREA 2     Internal room or area flooding of a HU3.2  Internal room or area FLOODING    Replaced the term needed with ..required by technical magnitude sufficient to require            of a magnitude sufficient to      specifications This provides clarity for what it intended as being manual or automatic electrical             require manual or automatic       needed.
isolation of a SAFETY SYSTEM               electrical isolation of a SAFETY component needed for the current           SYSTEM component required by operating mode.                            technical specifications for the current operating mode 3     Movement of personnel within the    HU3.3  Movement of personnel within the  Replaced the phrase due to an offsite event to due to an PROTECTED AREA is impeded                  PROTECTED AREA is                 event external to the plant PROTECTED AREA The impact of a due to an offsite event involving          IMPEDED due to an event           hazardous material originating from offsite (outside the OCA) would hazardous materials (e.g., an              external to the PROTECTED         be the same as one originating from onsite but outside the plant offsite chemical spill or toxic gas        AREA involving hazardous          Protected Area.
release).                                  materials (e.g., an offsite chemical spill or toxic gas release) 4     A hazardous event that results in   HU3.4  A hazardous event that results in Added reference to Note 7.
on-site conditions sufficient to           on-site conditions sufficient to prohibit the plant staff from              prohibit the plant staff from accessing the site via personal            accessing the site via personal vehicles.                                  vehicles (Note 7) 5     (Site-specific list of natural or    N/A   N/A                               No other site-specific hazard has been identified for PSL.
technological hazard events) 76 of 124

EAL Comparison Matrix Note    EAL #3 does not apply to routine N/A Note 7: This EAL does not        This note, designated Note #7, is intended to apply to generic traffic impediments such as fog,             apply to routine traffic example EAL #4, not #3 as specified in the generic guidance.
snow, ice, or vehicle breakdowns             impediments such as or accidents.                                fog, snow, ice, or vehicle breakdowns or accidents.
77 of 124

EAL Comparison Matrix PSL NEI IC#              NEI IC Wording                              PSL IC Wording                              Difference/Deviation Justification IC#(s)
HU4    FIRE potentially degrading the         HU4    FIRE potentially degrading the     None level of safety of the plant.                 level of safety of the plant MODE: All                                     MODE: All NEI Ex.                                        PSL NEI Example EAL Wording                          PSL EAL Wording                               Difference/Deviation Justification EAL #                                         EAL #
1     a. A FIRE is NOT extinguished         HU4.1  A FIRE is not extinguished within   Table H-1 provides a list of site-specific fire areas.
within 15-minutes of ANY of the              15 min. of any of the following following FIRE detection                     FIRE detection indications (Note indications:                                 1):
Report from the field (i.e.,
Report from the field (i.e.,
visual observation) visual observation)
Receipt of multiple (more Receipt of multiple (more                   than 1) fire zone alarms or than 1) fire alarms or                      indications indications                               Field verification of a single fire zone alarm Field verification of a single            AND fire alarm                            The FIRE is located within any AND                                      Table H-1 area
: b. The FIRE is located within ANY of the following plant rooms or areas:
(site-specific list of plant rooms or areas) 2     a. Receipt of a single fire alarm     HU4.2  Receipt of a single fire zone       Table H-1 provides a list of site-specific fire areas.
(i.e., no other indications of a             alarm (i.e., no other indications Added Note 14 to clarify validation of a single fire zone alarm in FIRE).                                       of a FIRE) containment.
AND AND The fire zone alarm is indicating   Added justification to HU4.2 bases.
: b. The FIRE is located within                a FIRE within any Table H-1 Added (proved or disproved) to clarify the meaning of a 78 of 124

EAL Comparison Matrix ANY of the following plant rooms            area (excluding containment)         verified fire.
or areas:                                   (Note 14)
AND (site-specific list of plant rooms or The existence of a FIRE is not areas) verified (proved or disproved)
AND                                      within 30 min. of alarm receipt (Note 1)
: c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.
3     A FIRE within the plant or ISFSI      HU4.3 A FIRE within the PROTECTED          PSL has an ISFSI located inside the PSL plant Protected Area.
[for plants with an ISFSI outside           AREA not extinguished within 60 the plant Protected Area]                   min. of the initial report, alarm or PROTECTED AREA not                          indication (Note 1) extinguished within 60-minutes of the initial report, alarm or indication.
4     A FIRE within the plant or ISFSI      HU4.4 A FIRE within the PROTECTED          PSL has an ISFSI located inside the PSL plant Protected Area.
[for plants with an ISFSI outside           AREA that requires firefighting the plant Protected Area]                   support by an offsite fire PROTECTED AREA that requires                response agency to extinguish firefighting support by an offsite fire response agency to extinguish.
Note   Note: The Emergency Director           N/A  Note 1: The EC should declare        The classification timeliness note has been standardized across the should declare the Unusual Event                     the event promptly upon     PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                       determining that the time   EAL wording.
the applicable time has been                         limit has been exceeded, exceeded, or will likely be                          or will likely be exceeded.
exceeded.
Note                     N/A                  N/A  Note 14: A containment fire zone     Added Note 14 to clarify validation of a single fire zone alarm in alarm is considered       containment.
VALID upon receipt of multiple zones (more than one) actuated on CP-376 panel.
79 of 124

EAL Comparison Matrix Table H-1 PSL Fire Areas
* Reactor Containment Building/Shield Building
* Control Room
* Reactor Auxiliary Building (RAB)
* Diesel Generator Building
* Component Cooling Water (CCW)
* Intake Structure
* Intake Cooling Water (ICW)
* Steam Trestle Area
* Yard Areas:
o RWT o CST o DOST 80 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                               PSL IC Wording                   Difference/Deviation Justification IC#(s)
HU7    Other conditions exist which in the     HU7   Other conditions existing that in    None judgment of the Emergency                     the judgment of the EC warrant Director warrant declaration of a             declaration of a UE (NO)UE                                        MODE: All MODE: All NEI Ex.                                         PSL NEI Example EAL Wording                           PSL EAL Wording                   Difference/Deviation Justification EAL #                                          EAL #
1     Other conditions exist which in the    HU7.1  Other conditions exist which, in     None judgment of the Emergency                     the judgment of the EC, indicate Director indicate that events are in          that events are in progress or progress or have occurred which               have occurred which indicate a indicate a potential degradation of           potential degradation of the level the level of safety of the plant or           of safety of the plant or indicate a indicate a security threat to facility        security threat to facility protection has been initiated. No             protection has been initiated. No releases of radioactive material              releases of radioactive material requiring offsite response or                 requiring offsite response or monitoring are expected unless                monitoring are expected unless further degradation of safety                 further degradation of SAFETY systems occurs.                               SYSTEMS occurs.
81 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                            PSL IC Wording                            Difference/Deviation Justification IC#(s)
HA1    HOSTILE ACTION within the            HA1   HOSTILE ACTION within the           None OWNER CONTROLLED AREA                      OWNER CONTROLLED AREA or or airborne attack threat within 30        airborne attack threat within 30 minutes.                                   minutes MODE: All                                  MODE: All NEI Ex.                                        PSL NEI Example EAL Wording                          PSL EAL Wording                           Difference/Deviation Justification EAL #                                        EAL #
1    A HOSTILE ACTION is occurring or      HA1.1   A HOSTILE ACTION is              Example EALs #1 and #2 have been combined into a single EAL has occurred within the OWNER                 occurring or has occurred within for ease of use.
CONTROLLED AREA as reported                   the OWNER CONTROLLED by the (site-specific security shift          AREA as reported by PSL supervision).                                 Security Shift Supervision OR 2    A validated notification from NRC of an aircraft attack threat within 30           A validated notification from minutes of the site.                          NRC of an aircraft attack threat within 30 min. of the site 82 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                              PSL IC Wording                                  Difference/Deviation Justification IC#(s)
HA5    Gaseous release impeding            HA5   Gaseous release IMPEDING                Limited mode applicability to the modes specified in Table H-2.
access to equipment necessary             access to equipment necessary for normal plant operations,              for normal plant operations, cooldown or shutdown.                     cooldown or shutdown MODE: All                                 MODE: All (except RA3.2:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown only)
NEI Ex.                                     PSL NEI Example EAL Wording                        PSL EAL Wording                                   Difference/Deviation Justification EAL #                                      EAL #
1     a. Release of a toxic,             HA5.1  Release of a toxic, corrosive,         The site-specific list of plant rooms or areas with entry-related corrosive, asphyxiant or                  asphyxiant or flammable gas into       mode applicability are tabularized in Table H-2.
flammable gas into any of the             any Table H-2 room or area following plant rooms or areas:
AND (site-specific list of plant rooms Entry into the room or area is or areas with entry-related prohibited or IMPEDED (Note 5) mode applicability identified)
AND
: b. Entry into the room or area is prohibited or impeded.
Note   Note: If the equipment in the       N/A   Note 5: If the equipment in the listed None listed room or area was already                   room or area was already inoperable or out-of-service                      inoperable or out-of-service before the event occurred, then                   before the event occurred, then no emergency no emergency classification is                    classification is warranted.
warranted.
N/A     N/A                                N/A   Note 15: The Intake Structure          Note 15 provides clarification of what constitutes the Intake Structure for includes the Intake Cooling  the purpose of this EAL.
Waiter (ICW) system and 83 of 124

EAL Comparison Matrix those immediately surrounding areas that could potentially affect operability or access to ICW piping and components needed for safe shutdown.
Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area                           Mode(s)
Reactor Auxiliary Building (0.5 and 19.5 levels)              1, 2, 3, 4, 5 Turbine Building (east end levels required for access to AFW, 1, 2, 3, 4, 5 Steam Trestle, Switchgear and Control Rooms)
Cable Spreading Room                                          1, 2, 3, 4, 5 A3/B3 Switchgear Rooms                                        1, 2, 3, 4, 5 Steam Trestle Area                                            1, 2, 3, 4, 5 Intake Structure (Note 15)                                    1, 2, 3, 4, 5 Component Cooling Water Area                                  1, 2, 3, 4, 5 Emergency Diesel Generator Buildings                          1, 2, 3, 4, 5 Refueling Water Tank                                          1, 2, 3, 4, 5 Water Treatment Plant/Storage Tanks                            1, 2, 3, 4 Containment                                                       4, 5 84 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                           PSL IC Wording                            Difference/Deviation Justification IC#(s)
HA6    Control Room evacuation            HA6   Control Room evacuation          None resulting in transfer of plant           resulting in transfer of plant control to alternate locations.          control to alternate locations.
MODE: All                                MODE: All NEI Ex.                                     PSL NEI Example EAL Wording                       PSL EAL Wording                           Difference/Deviation Justification EAL #                                     EAL #
1    An event has resulted in plant     HA6.1  An event has resulted in the    Revised the generic wording to read: An event has resulted in the control being transferred from the        Control Room being evacuated    Control Room being evacuated. The phrase plant control being Control Room to (site-specific                                            transferred from the Control Room to (site-specific remote remote shutdown panels and                                                shutdown panels and local control stations). Is ambiguous and not local control stations).                                                  clearly defined within the remote shutdown procedures. The revised statement is clear and unambiguous.
The Control Room is deemed to have been evacuated when the last licensed operator leaves the Control Room.
85 of 124

EAL Comparison Matrix PSL NEI IC#               NEI IC Wording                                     PSL IC Wording                   Difference/Deviation Justification IC#(s)
HA7   Other conditions exist which in the         HA7    Other conditions exist that in the        None judgment of the Emergency Director                 judgment of the EC warrant declaration warrant declaration of an Alert.                   of an Alert MODE: All                                          MODE: All NEI Ex.                                             PSL NEI Example EAL Wording                                 PSL EAL Wording                   Difference/Deviation Justification EAL #                                             EAL #
1    Other conditions exist which, in the       HA7.1  Other conditions exist which, in the       None judgment of the Emergency Director,               judgment of the EC, indicate that events indicate that events are in progress or           are in progress or have occurred which have occurred which involve an actual or          involve an actual or potential substantial potential substantial degradation of the          degradation of the level of safety of the level of safety of the plant or a security        plant or a security event that involves event that involves probable life                 probable life threatening risk to site threatening risk to site personnel or             personnel or damage to site equipment damage to site equipment because of               because of HOSTILE ACTION. Any HOSTILE ACTION. Any releases are                  releases are expected to be limited to expected to be limited to small fractions         small fractions of the EPA Protective of the EPA Protective Action Guideline            Action Guideline exposure levels.
exposure levels.
86 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                                PSL IC Wording                  Difference/Deviation Justification IC#(s)
HS1    HOSTILE ACTION within the            HS1    HOSTILE ACTION within the            None PROTECTED AREA                              PROTECTED AREA MODE: All                                   MODE: All NEI Ex.                                       PSL NEI Example EAL Wording                            PSL EAL Wording                 Difference/Deviation Justification EAL #                                       EAL #
1    A HOSTILE ACTION is occurring        HS1.1  A HOSTILE ACTION is occurring or has None or has occurred within the                  occurred within the PROTECTED AREA PROTECTED AREA as reported                  as reported by PSL Security Shift by the (site-specific security shift        Supervision supervision).
87 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                               PSL IC Wording                              Difference/Deviation Justification IC#(s)
HS6   Inability to control a key safety  HS6   Inability to control a key safety function  Deleted defueled mode applicability. Control of the cited function from outside the Control        from outside the Control Room               safety functions are not critical for a defueled reactor as there Room.                                                                                is no energy source in the RPV or RCS.
MODE: 1 - Power Operation, 2 - Startup, MODE: All                                3 - Hot Standby, 4 - Hot Shutdown, 5 -      This revised mode applicability is a deviation from the Cold Shutdown, 6 - Refueling                NEI 99-01 Revision 6 HS6 generic guidance but is deemed acceptable consistent with endorsed NRC EP FAQ 2015-014.
NEI Ex.                                    PSL NEI Example EAL Wording                            PSL EAL Wording                             Difference/Deviation Justification EAL #                                    EAL #
1    a. An event has resulted in       HS6.1  An event has resulted in the Control       Revised the wording of the first condition to read: An event plant control being transferred          Room being evacuated                       has resulted in the Control Room being evacuated. The from the Control Room to (site-                                                     phrase plant control being transferred from the Control AND specific remote shutdown panels                                                     Room to (site-specific remote shutdown panels and local Control of any of the following key and local control stations).                                                        control stations) is ambiguous and not clearly defined safety functions is not re-established within the remote shutdown procedures. The revised AND                                 within 15 min. of the last licensed statement is clear and unambiguous.
operator leaving the Control Room
: b. Control of ANY of the (Note 1):                                  The Hot Shutdown Control Panel is the site-specific remote following key safety functions is Reactivity (Modes 1, 2 and 3 only)    shutdown panels and local control stations.
not reestablished within (site-Core Cooling specific number of minutes).                                                        Added the words of the last licensed operator leaving the RCS heat removal Control Room to provide criteria for when the 15 minutes Reactivity control control clock begins.
Core cooling [PWR] / RPV The Mode applicability for the reactivity control safety water level [BWR]
function has been limited to Modes 1, 2, and 3. In Modes 4, RCS heat removal                                                                 5 and 6, adequate shutdown margin exists under all conditions.
This revised mode applicability is a deviation from the NEI 99-01 Revision 6 HS6 generic guidance but is deemed acceptable consistent with endorsed NRC EP 88 of 124

EAL Comparison Matrix FAQ 2015-014.
89 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                                  PSL IC Wording                  Difference/Deviation Justification IC#(s)
HS7    Other conditions exist which in      HS7   Other conditions existing that in the      None the judgment of the Emergency              judgment of the EC warrant declaration Director warrant declaration of a          of a Site Area Emergency Site Area Emergency.                       MODE: All MODE: All NEI Ex.                                      PSL NEI Example EAL Wording                              PSL EAL Wording                   Difference/Deviation Justification EAL #                                       EAL #
1     Other conditions exist which in     HS7.1  Other conditions exist which, in the       None the judgment of the Emergency              judgment of the EC, indicate that events Director indicate that events are          are in progress or have occurred which in progress or have occurred               involve actual or likely major failures of which involve actual or likely             plant functions needed for protection of major failures of plant functions          the public or HOSTILE ACTION that needed for protection of the               results in intentional damage or malicious public or HOSTILE ACTION that              acts, (1) toward site personnel or results in intentional damage or           equipment that could lead to the likely malicious acts, (1) toward site            failure of or, (2) that prevent effective personnel or equipment that                access to equipment needed for the could lead to the likely failure of        protection of the public. Any releases are or, (2) that prevent effective             not expected to result in exposure levels access to equipment needed for             which exceed EPA Protective Action the protection of the public. Any          Guideline exposure levels beyond the releases are not expected to               SITE BOUNDARY.
result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
90 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                     PSL IC Wording                   Difference/Deviation Justification IC#(s)
HG1   HOSTILE ACTION resulting in          N/A   N/A                 IC HG1 and associated example EAL are not implemented in loss of physical control of the                                the PSL scheme.
facility.
There are several other ICs that are redundant with this IC, MODE: All                                                      and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA                                                                        051, clarified the intended emergency classification level for spent fuel pool level events.
This exclusion of the generic HG1 guidance is a deviation from the NEI 99-01 Revision generic guidance but is deemed acceptable consistent with endorsed NRC EP FAQ 2015-013.
NEI Ex.                                      PSL NEI Example EAL Wording                   PSL EAL Wording                  Difference/Deviation Justification EAL #                                      EAL #
1    a. A HOSTILE ACTION is               N/A   N/A                  IC HG1 and associated example EAL is not implemented in occurring or has occurred within                                the PSL scheme.
the PROTECTED AREA as There are several other ICs that are redundant with this IC, reported by the (site-specific and are better suited to ensure timely and effective security shift supervision).
emergency declarations. In addition, the development of new AND                                                         spent fuel pool level EALs, as a result of NRC Order EA                                                                         051, clarified the intended emergency classification level for
: b. EITHER of the following has spent fuel pool level events. This deviation is justified occurred:
because:
: 1. ANY of the following safety
: 1. Hostile Action in the Protected Area is bounded by ICs functions cannot be HS1 and HS7. Hostile Action resulting in a loss of controlled or maintained.
physical control is bound by EAL HG7, as well as any Reactivity control                                    event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA)
Core cooling                                          Protective Action Guides (PAGs).
[PWR]/RPV water level [BWR]                                           a. If, for whatever reason, the Control Room must be 91 of 124

EAL Comparison Matrix RCS heat removal                               evacuated, and control of safety functions (e.g.,
reactivity control, core cooling, and RCS heat OR removal) cannot be reestablished, then IC HS6 would
: 2. Damage to spent fuel has                            apply, as well as IC HS7 if desired by the EAL occurred or is IMMINENT.                            decision-maker.
: b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG7.
: c. From a Hostile Action perspective, ICs HS1, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
: d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
: 2. Any event which causes a loss of spent fuel pool level will be bounded by ICs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.
: a. An event that leads to a radiological release will be bounded by ICs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs AG1 and HG7, thus making this part of HG1 redundant and unnecessary.
ICs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG7 have been implemented consistent with NEI 99-01 Revision 6 and thus HG1 is adequately bounded as described above.
This exclusion of the generic HG1 guidance is a deviation from the NEI 99-01 Revision generic guidance but is deemed acceptable consistent with endorsed NRC EP FAQ 2015-013.
NEI IC#          NEI IC Wording        PSL PSL IC Wording                Difference/Deviation Justification 92 of 124

EAL Comparison Matrix IC#(s)
HG7    Other conditions exist which in    HG7    Other conditions exist which in the        None the judgment of the Emergency             judgment of the EC warrant declaration Director warrant declaration of a         of a General Emergency General Emergency                         MODE: All MODE: All NEI Ex.                                    PSL NEI Example EAL Wording                           PSL EAL Wording                     Difference/Deviation Justification EAL #                                     EAL #
1     Other conditions exist which in   HG7.1  Other conditions exist which, in the        None the judgment of the Emergency            judgment of the EC, indicate that events Director indicate that events are        are in progress or have occurred which in progress or have occurred             involve actual or IMMINENT substantial which involve actual or                  core degradation or melting with IMMINENT substantial core                potential for loss of containment integrity degradation or melting with              or HOSTILE ACTION that results in an potential for loss of containment        actual loss of physical control of the integrity or HOSTILE ACTION              facility. Releases can be reasonably that results in an actual loss of        expected to exceed EPA Protective physical control of the facility.        Action Guideline exposure levels offsite Releases can be reasonably               for more than the immediate site area.
expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
93 of 124

EAL Comparison Matrix Category S System Malfunction 94 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                         PSL IC Wording                                Difference/Deviation Justification IC#(s)
SU1   Loss of all offsite AC power       SU1   Loss of all offsite AC power        None capability to emergency buses for        capability to emergency buses for 15 minutes or longer.                    15 minutes or longer MODE: Power Operation, Startup,          MODE: 1 - Power Operation, Hot Standby, Hot Shutdown                2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown NEI Ex.                                    PSL NEI Example EAL Wording                     PSL EAL Wording                                 Difference/Deviation Justification EAL #                                    EAL #
1    Loss of ALL offsite AC power      SU1.1  Loss of all offsite AC power        4.16 kV emergency buses 1[2]A3 and 1[2]B3 are the site-specific capability to (site-specific             capability, Table 1[2]S-1, to 4.16  emergency buses.
emergency buses) for 15 minutes          kV emergency buses 1[2]A3 and       Table S-1 lists credited offsite 4.16 kV emergency bus AC power or longer.                               1[2]B3 for  15 min. (Note 1)       sources.
Note   The Emergency Director should      N/A   Note 1: The EC should declare       The classification timeliness note has been standardized across the declare the Unusual Event                          the event promptly upon   PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that 15                  determining that the time EAL wording.
minutes has been exceeded, or                      limit has been exceeded, will likely be exceeded.                           or will likely be exceeded.
95 of 124

EAL Comparison Matrix Table 1[2]S-1 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer Onsite
* 1A[2A] Auxiliary Transformer
* 1B[2B] Auxiliary Transformer
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned) 96 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                         PSL IC Wording                               Difference/Deviation Justification IC#(s)
SU2    UNPLANNED loss of Control         SU3   UNPLANNED loss of Control           None Room indications for 15 minutes         Room indications for 15 minutes or longer.                              or longer.
MODE: Power Operation,                  MODE: 1 - Power Operation, Startup, Hot Standby, Hot               2 - Startup, 3 - Hot Standby, Shutdown                                4 - Hot Shutdown NEI Ex.                                   PSL NEI Example EAL Wording                      PSL EAL Wording                               Difference/Deviation Justification EAL #                                    EAL #
1     An UNPLANNED event results in    SU3.1  An UNPLANNED event results in       The site-specific Safety System Parameter list is tabulated in Table the inability to monitor one or         the inability to monitor one or     S-2.
more of the following parameters        more Table S-2 parameters from from within the Control Room for        within the Control Room for  15 15 minutes or longer.                   min. (Note 1)
Note   The Emergency Director should     N/A   Note 1: The EC should declare       The classification timeliness note has been standardized across the declare the Unusual Event                         the event promptly upon   PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                    determining that the time EAL wording.
15 minutes has been exceeded,                     limit has been exceeded, or will likely be exceeded.                       or will likely be exceeded.
97 of 124

EAL Comparison Matrix
[BWR parameter list]                         [PWR parameter list]
Reactor Power                               Reactor Power RPV Water Level                             RCS Level RPV Pressure                                RCS Pressure Primary Containment Pressure                In-Core/Core Exit Temperature Suppression Pool Level                      Levels in at least (site-specific number) steam generators Suppression Pool Temperature                Steam Generator Auxiliary or Emergency Feed Water Flow Table S-2       Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CET temperature
* Level in at least one SG
* Auxiliary feedwater flow to at least one SG 98 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                               PSL IC Wording                              Difference/Deviation Justification IC#(s)
SU3    Reactor coolant activity greater     SU4      Reactor coolant activity greater   None than Technical Specification                  than Technical Specification allowable limits.                             allowable limits MODE: Power Operation, Startup,               MODE: 1 - Power Operation, Hot Standby, Hot Shutdown                     2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown NEI Ex.                                      PSL NEI Example EAL Wording                         PSL EAL Wording                              Difference/Deviation Justification EAL #                                      EAL #
1    (Site-specific radiation monitor)    N/A    N/A                                 There are no site-specific radiation monitor readings that indicative reading greater than (site-specific                                             of RCS activity > Technical Specification limits.
value).
2    Sample analysis indicates that a    SU4.1   Sample analysis indicates that a    PSL Technical Specification 3.4.8 RCS Specific Activity provides reactor coolant activity value is           reactor coolant activity value is   the Technical Specification allowable coolant activity limits.
greater than an allowable limit             > any of the following allowable specified in Technical                      limits:
Specifications.
* 60 Ci/gm dose equivalent I-131 instantaneous
* 1.0 Ci/gm dose equivalent I-131 for > 48 hours
* 518.9 Ci/gm dose equivalent Xe-133 for > 48 hours 99 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                         PSL IC Wording                             Difference/Deviation Justification IC#(s)
SU4    RCS leakage for 15 minutes or     SU5   RCS leakage for 15 minutes or     None longer.                                 longer MODE: Power Operation,                  MODE: 1 - Power Operation, Startup, Hot Standby, Hot               2 - Startup, 3 - Hot Standby, Shutdown                                4 - Hot Shutdown NEI Ex.                                    PSL NEI Example EAL Wording                      PSL EAL Wording                             Difference/Deviation Justification EAL #                                    EAL #
1    RCS unidentified or pressure      SU5.1  RCS unidentified or pressure      Example EALs #1, 2 and 3 have been combined into a single EAL boundary leakage greater than            boundary leakage > 10 gpm for     for usability.
(site-specific value) for 15              15 min.
minutes or longer.                         OR RCS identified leakage > 25 gpm 2    RCS identified leakage greater           for  15 min.
than (site-specific value) for 15          OR minutes or longer.                       Leakage from the RCS to a location outside containment 3    Leakage from the RCS to a                > 25 gpm for  15 min.
location outside containment             (Note 1) greater than 25 gpm for 15 minutes or longer.
Note   The Emergency Director should      N/A   Note 1: The EC should declare     The classification timeliness note has been standardized across declare the Unusual Event                         the event promptly upon  the PSL EAL scheme by referencing the "time limit" specified promptly upon determining that                    determining that the     within the EAL wording.
15 minutes has been exceeded,                     time limit has been or will likely be exceeded.                       exceeded, or will likely be exceeded.
100 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                           PSL IC Wording                                  Difference/Deviation Justification IC#(s)
SU5     Automatic or manual (trip          SU6   Automatic or manual trip fails to   Included Mode 2 - Startup applicability. Reactivity control Safety
[PWR] / scram [BWR]) fails to            shut down the reactor               Function Status Check acceptance criteria includes reactor power
                                                                                                 -4 shutdown the reactor.                    MODE: 1 - Power Operation,          < 5 x 10 %. Mode 1 - Power Operations is defined as reactor power MODE: Power Operation                    2 - Startup                         > 5%.
NEI Ex.                                      PSL NEI Example EAL Wording                        PSL EAL Wording                                Difference/Deviation Justification EAL #                                       EAL #
1      a. An automatic (trip [PWR] /     SU6.1  An automatic trip did not shut     As specified in the generic developers guidance Developers may scram [BWR]) did not shutdown            down the reactor as indicated by   include site-specific EOP criteria indicative of a successful reactor the reactor.                             reactivity control Safety Function shutdown in an EAL statement, the Basis or both (e.g., a reactor Status Check acceptance criteria   power level). Consistent with the CEOG Emergency Procedure AND                                   not met after any RPS setpoint     Guidelines, a successful shutdown is defined by Safety Function
: b. A subsequent manual action            is exceeded                        Status Check acceptance criteria being met.
taken at the reactor control               AND                           Added the words "... after any RPS setpoint is exceeded" to clarify consoles is successful in A subsequent automatic trip or     that it is a failure of the automatic trip when a valid trip signal has shutting down the reactor.
manual trip (RX Trip               been exceeded.
Pushbuttons) are successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria met (Note 8) 2     a. A manual trip ([PWR] /          SU6.2  A manual trip (RX Trip             As specified in the generic developers guidance Developers may scram [BWR]) did not shutdown             Pushbuttons) did not shut down     include site-specific EOP criteria indicative of a successful reactor the reactor.                              the reactor as indicated by        shutdown in an EAL statement, the Basis or both (e.g., a reactor reactivity control Safety Function power level). Consistent with the CEOG Emergency Procedure AND                                   Status Check acceptance criteria   Guidelines, a successful shutdown is defined by Safety Function
: b. EITHER of the following:              not met (Note 8)                   Status Check acceptance criteria being met.
: 1. A subsequent manual                AND action taken at the              A subsequent automatic trip or reactor control consoles         manual trip action (RX Trip is successful in shutting        Pushbuttons) is successful in 101 of 124

EAL Comparison Matrix down the reactor.              shutting down the reactor as indicated by reactivity control OR Safety Function Status Check 2 A subsequent automatic         acceptance criteria met (trip [PWR] / scram
[BWR]) is successful in shutting down the reactor.
Note: A manual action is any Notes                                     N/A Note 8: A manual action is any      None operator action, or set of operator action, or set of actions, which causes the actions, which causes control rods to be rapidly the control rods to be inserted into the core, and does rapidly inserted into the not include manually driving in core, and does not control rods or implementation include manually driving of boron injection strategies.
in control rods or implementation of boron injection strategies.
102 of 124

EAL Comparison Matrix PSL NEI IC#           NEI IC Wording                        PSL IC Wording                            Difference/Deviation Justification IC#(s)
SU6    Loss of all onsite or offsite  SU7   Loss of all onsite or offsite   None communications capabilities.         communications capabilities.
MODE: Power Operation,               MODE: 1 - Power Operation, Startup, Hot Standby, Hot            2 - Startup, 3 - Hot Standby, Shutdown                             4 - Hot Shutdown NEI Ex.                                PSL NEI Example EAL Wording                   PSL EAL Wording                            Difference/Deviation Justification EAL #                                 EAL #
1     Loss of ALL of the following  SU7.1  Loss of all Table S-4 onsite    Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods:        communication methods           for simplification of presentation.
(site-specific list of                   OR                          Table S-4 provides a site-specific list of onsite, State and local communications methods)              Loss of all Table S-4 State and agency (ORO) and NRC communications methods.
local agency communication 2     Loss of ALL of the following         methods ORO communications methods:              OR (site-specific list of               Loss of all Table S-4 NRC communications methods) communication methods 3     Loss of ALL of the following NRC communications methods:
(site-specific list of communications methods) 103 of 124

EAL Comparison Matrix Table S-4 Communication Methods State/
System                      Onsite        NRC Local Plant Radio System                                   X Commercial Telephone System                          X      X     X Plant Address (PA) System                            X Federal Telephone System (ENS)                                    X EMnet                                                       X Hot Ring Down (HRD) Telephone System                        X Shift Manager Cell Phones                                   X     X 104 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                           PSL IC Wording                              Difference/Deviation Justification IC#(s)
SU7    Failure to isolate containment or  SU8   Failure to isolate containment or   None loss of containment pressure             loss of containment pressure control. [PWR]                           control MODE: Power Operation,                   MODE: 1 - Power Operation, Startup, Hot Standby, Hot                2 - Startup, 3 - Hot Standby, Shutdown                                 4 - Hot Shutdown NEI Ex.                                     PSL NEI Example EAL Wording                      PSL EAL Wording                               Difference/Deviation Justification EAL #                                     EAL #
1     a. Failure of containment to      SU8.1  Any penetration is not closed       Example EALs #1 and #2 have been combined for usability.
isolate when required by an              within 15 min. of a VALID Containment isolation actuation is initiated by a CIS[CIAS]
actuation signal.                        CIS[CIAS] actuation signal actuation signal.
OR AND CTMT pressure > 10 psig [5.4        Containment pressure greater than 10 psia [5.4 psig] is the
: b. ALL required penetrations             psig] with < one full train of CTMT pressure at which containment heat removal equipment are are not closed within 15 minutes         heat removal systems (Note 11)      designed to automatically actuate.
of the actuation signal.                 operating per design for  15 min.
(Note 1) 2     a. Containment pressure greater than (site-specific pressure).
AND
: b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.
N/A    N/A                                N/A   Note 1: The EC should declare       Added note 1 consistent with other EALs with a timing component.
the event promptly upon determining that the time 105 of 124

EAL Comparison Matrix limit has been exceeded, or will likely be exceeded.
N/A    N/A          N/A Note 11: One full train of         Added note 11 to clarify what constitutes a full train of containment containment heat         heat removal systems.
removal systems consist of any one of the following:
* Two (2)
Containment Spray Headers with flow in each header at least 2,550 [2,700] gpm
* Four (4)
Containment Coolers
* Two (2)
Containment Coolers with at least one (1) Containment Spray Header with at least 2,550
[2,700] gpm 106 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                              PSL IC Wording                                 Difference/Deviation Justification IC#(s)
SA1    Loss of all but one AC power          SA1   Loss of all but one AC power          None source to emergency buses for               source to emergency buses for 15 minutes or longer.                       15 minutes or longer MODE: Power Operation,                      MODE: 1 - Power Operation, Startup, Hot Standby, Hot                   2 - Startup, 3 - Hot Standby, Shutdown                                    4 - Hot Shutdown NEI Ex.                                       PSL NEI Example EAL Wording                         PSL EAL Wording                                  Difference/Deviation Justification EAL #                                        EAL #
1     a. AC power capability to (site-     SA1.1  AC power capability, Table S-1,       4.16 kV emergency buses 1[2]A3 and 1[2]B3 are the site-specific specific emergency buses) is                to 4.16 kV emergency buses            emergency buses.
reduced to a single power source            1[2]A3 and 1[2]B3 reduced to a        Table S-1 lists credited offsite and onsite 4.16 kV emergency bus AC for 15 minutes or longer.                   single power source for               power sources.
15 min. (Note 1)
AND AND
: b. Any additional single power              Any additional single power source failure will result in a loss        source failure will result in loss of of all AC power to SAFETY                   all AC power to SAFETY SYSTEMS.                                    SYSTEMS Note   The Emergency Director should         N/A   Note 1: The EC should declare         The classification timeliness note has been standardized across the declare the Alert promptly upon                     the event promptly upon       PSL EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has                     determining that the time     EAL wording.
been exceeded, or will likely be                    limit has been exceeded, exceeded.                                           or will likely be exceeded.
107 of 124

EAL Comparison Matrix Table 1[2]S-1 Unit 1[2] AC Power Sources Offsite
* 1A[2A] Startup Transformer
* 1B[2B] Startup Transformer Onsite
* 1A[2A] Auxiliary Transformer
* 1B[2B] Auxiliary Transformer
* 1A[2A] Diesel Generator
* 1B[2B] Diesel Generator
* 1AB - 2AB SBO Cross Tie
* 2A4 (2B4) Cross Tie (if already aligned) 108 of 124

EAL Comparison Matrix PSL NEI IC#            NEI IC Wording                           PSL IC Wording                            Difference/Deviation Justification IC#(s)
SA2    UNPLANNED loss of Control          SA3   UNPLANNED loss of Control        None Room indications for 15 minutes          Room indications for 15 minutes or longer with a significant             or longer with a significant transient in progress.                   transient in progress.
MODE: Power Operation,                   MODE: 1 - Power Operation, Startup, Hot Standby, Hot                2 - Startup, 3 - Hot Standby, Shutdown                                 4 - Hot Shutdown NEI Ex.                                     PSL NEI Example EAL Wording                      PSL EAL Wording                            Difference/Deviation Justification EAL #                                      EAL #
1      An UNPLANNED event results in     SA3.1 An UNPLANNED event results in    The site-specific Safety System Parameter list is in Table S-2.
the inability to monitor one or         the inability to monitor one or The significant transient list has been tabularized in Table S-3 for more of the following parameters        more Table S-2 parameters from ease of use.
from within the Control Room for        within the Control Room for  15 15 minutes or longer.                   min. (Note 1)
AND                                    AND ANY of the following transient           Any significant transient is in events in progress.                      progress, Table S-3 Automatic or manual runback greater than 25%
thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor scram [BWR] / trip
[PWR]
109 of 124

EAL Comparison Matrix ECCS (SI) actuation Thermal power oscillations greater than 10% [BWR]
Note    The Emergency Director should       N/A Note 1: The EC should declare     The classification timeliness note has been standardized across the declare the Unusual Event                       the event promptly upon   PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                  determining that the time EAL wording.
15 minutes has been exceeded,                   limit has been exceeded, or will likely be exceeded.                     or will likely be exceeded.
110 of 124

EAL Comparison Matrix
[BWR parameter list]                         [PWR parameter list]
Reactor Power                    Reactor Power RPV Water Level                  RCS Level RPV Pressure                    RCS Pressure Primary Containment Pressure     In-Core/Core Exit Temperature Suppression Pool Level           Levels in at least (site-specific number) steam generators Suppression Pool Temperature     Steam Generator Auxiliary or Emergency Feed Water Flow Table S-2         Safety System Parameters
* Reactor power
* RCS level
* RCS pressure
* CET temperature
* Level in at least one SG
* Auxiliary feedwater flow to at least one SG Table S-3         Significant Transients
* Electrical load rejection > 25% full electrical load
* Reactor Trip
* SIAS actuation 111 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                             PSL IC Wording                              Difference/Deviation Justification IC#(s)
SA5    Automatic or manual (trip [PWR]       SA6   Automatic or manual trip fails to  Included Mode 2 - Startup applicability. Reactivity control Safety
         / scram [BWR]) fails to shutdown            shut down the reactor and          Function Status Check acceptance criteria includes reactor power
                                                                                                -4 the reactor, and subsequent                 subsequent manual actions          < 5 x 10 %. Mode 1 - Power Operations is defined as reactor power manual actions taken at the                 taken at the reactor control       > 5%.
reactor control consoles are not            consoles are not successful in successful in shutting down the             shutting down the reactor reactor.                                    MODE: 1 - Power Operation, MODE: Power Operation                       2 - Startup NEI Ex.                                       PSL NEI Example EAL Wording                         PSL EAL Wording                              Difference/Deviation Justification EAL #                                        EAL #
1     a. An automatic or manual (trip     SA6.1  An automatic or manual trip (RX    As specified in the generic developers guidance Developers may
[PWR] / scram [BWR]) did not               Trip Pushbuttons) did not shut     include site-specific EOP criteria indicative of a successful reactor shutdown the reactor.                      down the reactor as indicated by   shutdown in an EAL statement, the Basis or both (e.g., a reactor reactivity control Safety Function power level). Consistent with the CEOG Emergency Procedure AND                                     Status Check acceptance criteria   Guidelines, a successful shutdown is defined by Safety Function
: b. Manual actions taken at the             not met                            Status Check acceptance criteria being met.
reactor control consoles are not                AND                           The reactor trip pushbuttons are the only means of initiating a successful in shutting down the Subsequent automatic or manual     manual trip from the reactor control consoles.
reactor.
trip actions (RX Trip Pushbuttons) are not successful in shutting down the reactor as indicated by reactivity control Safety Function Status Check acceptance criteria not met (Note 8)
Note: A manual action is any Notes                                        N/A   Note 8: A manual trip action is    None operator action, or set of actions, any operator action, or which causes the control rods to set of actions, which be rapidly inserted into the core, causes the control rods and does not include manually to be rapidly inserted driving in control rods or 112 of 124

EAL Comparison Matrix implementation of boron injection into the core, and does strategies.                       not include manually driving in control rods or implementation of boron injection strategies 113 of 124

EAL Comparison Matrix PSL NEI IC#           NEI IC Wording                     PSL IC Wording                         Difference/Deviation Justification IC#(s)
SA9    Hazardous event affecting a SA9.1  Hazardous event affecting      Revised wording from affecting a SAFETY SYSTEM to read SAFETY SYSTEM needed for           SAFETY SYSTEMS needed for      affecting SAFETY SYSTEMS to align with changes made the current operating mode.        the current operating mode     consistent with NRC EP FAQ 2016-002.
MODE: Power Operation,             MODE: 1 - Power Operation, Startup, Hot Standby, Hot          2 - Startup, 3 - Hot Standby, Shutdown                           4 - Hot Shutdown 114 of 124

EAL Comparison Matrix NEI Ex.                                   PSL NEI Example EAL Wording                     PSL EAL Wording                             Difference/Deviation Justification EAL #                                    EAL #
: a. The occurrence of ANY of      SA9.1                                The hazardous events have been tabularized in Table S-5.
1                                             The occurrence of any Table the following hazardous events:
S-5 hazardous event           The proposed PSL CA6.1 and SA9.1 wording is intended to ensure Seismic event (earthquake)
Internal or external flooding           AND                        that an Alert should be declared only when actual or potential event                                                             performance issues with SAFETY SYSTEMS have occurred as a Event damage has caused        result of a hazardous event. The occurrence of a hazardous event High winds or tornado strike indications of degraded        will result in an Unusual Event classification at a minimum. In order FIRE performance on one train of a  to warrant escalation to the Alert classification, the hazardous event EXPLOSION (site-specific hazards)
SAFETY SYSTEM needed for       should cause indications of degraded performance to one train of a Other events with similar the current operating mode     SAFETY SYSTEM with either indications of degraded performance hazard characteristics as              AND EITHER:                on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the determined by the Shift                                           second SAFETY SYSTEM train, such that the operability or Event damage has         reliability of the second train is a concern. In addition, escalation to Manager AND                                       caused indications of    the Alert classification should not occur if the damage from the
: b. EITHER of the following:                 degraded performance to  hazardous event is limited to a SAFETY SYSTEM that was
: 1. Event damage has                      the second train of the  inoperable, or out of service, prior to the event occurring. As such, caused indications of                SAFETY SYSTEM            the proposed EALs will reduce the potential of declaring an Alert degraded performance in              needed for the current   when events are in progress that do not involve an actual or at least one train of a              operating mode           potential substantial degradation of the level of safety of the plant, SAFETY SYSTEM                       Event damage has i.e., does not cause significant concern with shutting down or needed for the current               resulted in VISIBLE      cooling down the plant.
operating mode.                      DAMAGE to the second     EALs CA6.1 and SA9.1 do not directly escalate to a Site Area OR                                 train of the SAFETY      Emergency or a General Emergency due to a hazardous event.
: 2. The event has caused                  SYSTEM needed for the    The Fission Product Barrier and/or Abnormal Radiation VISIBLE DAMAGE to a                  current operating mode   Levels/Radiological Effluent recognition categories would provide SAFETY SYSTEM                                                 an escalation path to a Site Area Emergency or a General component or structure           (Notes 9, 10)
Emergency.
needed for the current operating mode.                                               The EALs and the Basis sections have been revised to ensure potential escalations from an Unusual Event to an Alert, due to a hazardous event, is appropriate as the concern with these EALs is:
(1) a hazardous event has occurred, (2) one SAFETY SYSTEM train is having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or the VISIBLE DAMAGE is enough to be concerned that the second SAFETY SYSTEM train may have 115 of 124

EAL Comparison Matrix operability or reliability issues.
The definition for VISIBLE DAMAGE has been revised to reflect the fact that the EALs are based upon SAFETY SYSTEM trains rather than individual components or structures.
Note 9 has been added to CA6.1 and SA9.1 as it meets the intent of the EALs, is consistent with other EALs (e.g., EAL HA5.1 which was previously endorsed by the NRC), and ensures that declared emergencies are based upon unplanned events with the potential to pose a radiological risk to the public.
Note 10 has been added to CA6.1 and SA9.1 to help reinforce and succinctly capture the more detailed information from the revised basis section related to when conditions would require the declaration of an Alert.
CA6.1 and SA9.1 are consistent with NRC FAQ 2016-002 requiring degraded performance or visible damage to more than one safety system train caused by the specified events.
This revised wording is a deviation from the NEI 99-01 Revision 6 CA6 and SA9 generic wording and bases but is deemed acceptable consistent with endorsed NRC EP FAQ 2016-002.
116 of 124

EAL Comparison Matrix N/A    N/A          N/A Note 9: If the affected SAFETY    Added Note 9 consistent with the recommendation of NRC EP FAQ SYSTEM train was         2016-002.
already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.
Note 10: If the hazardous event   Added Note 10 consistent with the recommendation of NRC EP only resulted in         FAQ 2016-002.
VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
117 of 124

EAL Comparison Matrix Table S-5      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 118 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                          PSL IC Wording                               Difference/Deviation Justification IC#(s)
SS1    Loss of all offsite and all onsite  SS1   Loss of all offsite power and all  None AC power to emergency buses               onsite AC power to emergency for 15 minutes or longer.                 buses for 15 minutes or longer MODE: Power Operation,                    MODE: 1 - Power Operation, Startup, Hot Standby, Hot                 2 - Startup, 3 - Hot Standby, Shutdown                                  4 - Hot Shutdown NEI Ex.                                     PSL NEI Example EAL Wording                      PSL EAL Wording                                Difference/Deviation Justification EAL #                                      EAL #
1     Loss of ALL offsite and ALL        SS1.1  Loss of all offsite and all onsite 4.16 kV emergency buses 1[2]A3 and 1[2]B3 are the site-specific onsite AC power to (site-specific         AC power to 4.16 kV                emergency buses.
emergency buses) for 15 minutes           emergency buses 1[2]A3 and or longer.                                1[2]B3 for  15 min. (Note 1)
Note   The Emergency Director should       N/A   Note 1: The EC should declare      The classification timeliness note has been standardized across the declare the Unusual Event                           the event promptly       PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                      upon determining that    EAL wording.
15 minutes has been exceeded,                       the time limit has been or will likely be exceeded.                         exceeded, or will likely be exceeded.
119 of 124

EAL Comparison Matrix PSL NEI IC#               NEI IC Wording                           PSL IC Wording                                 Difference/Deviation Justification IC#(s)
SS5      Inability to shutdown the reactor  SS6    Inability to shut down the         Included Mode 2 - Startup applicability. Reactivity control Safety causing a challenge to (core              reactor causing a challenge to     Function Status Check acceptance criteria includes reactor power
                                                                                                -4 cooling [PWR] / RPV water level           core cooling or RCS heat           < 5 x 10 %. Mode 1 - Power Operations is defined as reactor power
[BWR]) or RCS heat removal.               removal                            > 5%.
MODE: Power Operation                     MODE: 1 - Power Operation, 2 - Startup NEI Ex.                                       PSL NEI Example EAL Wording                        PSL EAL Wording                                Difference/Deviation Justification EAL #                                       EAL #
The time the EAL is assessed 1       a. An automatic or manual (trip   SS6.1  An automatic or manual trip did    As specified in the generic developers guidance Developers may
[PWR] / scram [BWR]) did not not shut down the reactor as       include site-specific EOP criteria indicative of a successful reactor shutdown the reactor.
indicated by reactivity control    shutdown in an EAL statement, the Basis or both (e.g., a reactor AND Safety Function Status Check       power level). Consistent with the CEOG Emergency Procedure
: b. All manual actions to acceptance criteria not met        Guidelines, a successful shutdown is defined by Safety Function shutdown the reactor have AND                             Status Check acceptance criteria being met.
been unsuccessful.
All actions taken to shut down AND                                                                      Added the word taken to the second condition to emphasize the the reactor are not successful
: c. EITHER of the following                                                  intent that it is all actions taken up to the point of either core as indicated by reactivity control conditions exist:                                                        cooling or heat sink is challenged are not successful and to not Safety Function Status Check
              * (Site-specific indication of                                           wait until all possible actions have been completed.
acceptance criteria not met an inability to adequately AND EITHER:                     Core Damage Assessment uses a CET temperature of 1200 &deg;F as the remove heat from the core) temperature indicative of a significantly damaged core.
              * (Site-specific indication of
* REP CETs >1200&deg;F an inability to adequately                                            Failure to meet the applicable HR acceptance criteria indicates an
* Applicable RCS and Core       extreme challenge to the ability to remove RCS heat using the remove heat from the RCS)                Heat Removal (HR) Safety     steam generators (i.e., loss of an effective secondary-side heat Function Status Check        sink).
acceptance criteria not met 120 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                        PSL IC Wording                                Difference/Deviation Justification IC#(s)
SS8    Loss of all Vital DC power for 15  SS2   Loss of all vital DC power for 15 None minutes or longer.                       minutes or longer.
MODE: Power Operation,                   MODE: 1 - Power Operation, Startup, Hot Standby, Hot                2 - Startup, 3 - Hot Standby, Shutdown                                 4 - Hot Shutdown NEI Ex.                                    PSL NEI Example EAL Wording                     PSL EAL Wording                               Difference/Deviation Justification EAL #                                     EAL #
1     Indicated voltage is less than    SS2.1  Indicated voltage < 112 VDC on    112 VDC is the site-specific minimum vital 125V DC bus voltage.
(site-specific bus voltage value)        both vital 125 VDC buses 1[2]A    Vital 125 VDC buses 1[2]A and 1[2]B are the site-specific vital DC on ALL (site-specific Vital DC           and 1[2]B for  15 min. (Note 1)  buses credited in this EAL.
busses) for 15 minutes or longer.
Note   The Emergency Director should      N/A   Note 1: The EC should declare     The classification timeliness note has been standardized across the declare the Unusual Event                        the event promptly upon   PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that                   determining that the      EAL wording.
15 minutes has been exceeded,                    time limit has been or will likely be exceeded.                      exceeded, or will likely be exceeded.
121 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                             PSL IC Wording                               Difference/Deviation Justification IC#(s)
SG1    Prolonged loss of all offsite and      SG1   Prolonged loss of all offsite and  None all onsite AC power to                       all onsite AC power to emergency buses.                             emergency buses MODE: Power Operation,                       MODE: 1 - Power Operation, Startup, Hot Standby, Hot                    2 - Startup, 3 - Hot Standby, Shutdown                                     4 - Hot Shutdown NEI Ex.                                         PSL NEI Example EAL Wording                           PSL EAL Wording                               Difference/Deviation Justification EAL #                                          EAL #
1     a. Loss of ALL offsite and ALL         SG1.1 Loss of all offsite and all onsite  4.16 kV emergency buses 1[2]A3 and 1[2]B3.
onsite AC power to (site-specific            AC power to 4.16 kV emergency       Core Damage Assessment uses a CET temperature of 1200 &deg;F as the emergency buses).                            buses 1[2]A3 and 1[2]B3             temperature indicative of a significantly damaged core.
AND                                         AND The proposed PSL SG1.1 omits the Station Blackout (SBO) coping
: b. EITHER of the following:                  REP CETs > 1200&#xba;F                   time threshold. As proposed, the General Emergency classification would be based on a loss of all onsite and offsite AC power to the
* Restoration of at least                                                 emergency buses with indications of degraded core cooling. The one AC emergency bus in                                                 PSL SBO analysis and derived coping time was determined in less than (site-specific                                                accordance with 10CFR50.63 and Regulatory Guide 1.155. This hours) is not likely.                                                   analysis does not take credit for plant capabilities in place to mitigate the effects of an extended loss of AC power
              *   (Site-specific indication of (ELAP). These capabilities were developed and implemented to an inability to adequately meet the requirements of NRC Orders EA-12-049 and EA-12-051, remove heat from the and pending regulations in 10 CFR 50.155 (per SECY-16-0142).
core)
In accordance with plant EOPs [1(2)-EOP-10], operators will declare an ELAP within 60 min. of the loss of all AC power to the emergency buses and direct implementation of FLEX Support Guidelines, including the deployment of dedicated portable equipment and performance of DC load shedding. Even if no AC emergency bus is energized, these actions will maintain or restore core cooling, containment, and spent fuel pool cooling capabilities indefinitely. Therefore, the underlying basis for the generic EAL coping time statement, that power must be restored to an AC emergency bus within a fixed amount of time to avoid a severe 122 of 124

EAL Comparison Matrix challenge to one or more fission product barriers, is not valid for PSL.
Additionally, the omission of the SBO coping time threshold does not remove the attribute of a likely General Emergency declaration prior to meeting the IC FG1 thresholds for ELAP events in which the RCS barrier has not been lost.
This revised wording is a deviation from the NEI 99-01 Revision 6 SG1 generic wording and bases but is deemed appropriate and acceptable.
Note    The Emergency Director should  N/A Note 1: The EC should declare    The classification timeliness note has been standardized across the declare the General Emergency              the event promptly       PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that             upon determining that    EAL wording.
(site-specific hours) has been             the time limit has been exceeded, or will likely be                exceeded, or will likely exceeded.                                  be exceeded.
123 of 124

EAL Comparison Matrix PSL NEI IC#             NEI IC Wording                         PSL IC Wording                               Difference/Deviation Justification IC#(s)
SG8    Loss of all AC and Vital DC        SG2   Loss of all emergency AC and       None power sources for 15 minutes or          vital DC power sources for 15 longer.                                  minutes or longer MODE: Power Operation,                   MODE: 1 - Power Operation, Startup, Hot Standby, Hot                2 - Startup, 3 - Hot Standby, Shutdown                                 4 - Hot Shutdown NEI Ex.                                     PSL NEI Example EAL Wording                     PSL EAL Wording                               Difference/Deviation Justification EAL #                                      EAL #
1     a. Loss of ALL offsite and ALL     SG2.1 Loss of all offsite and all onsite 4.16 kV emergency buses 1[2]A3 and 1[2]B3 are the site-specific onsite AC power to (site-specific        AC power to 4.16 kV                emergency buses.
emergency buses) for 15 minutes          emergency buses 1[2]A3 and         112 VDC is the site-specific minimum vital 125V DC bus voltage.
or longer.                               1[2]B3 for  15 min. (Note 1)
AND                            Vital 125 VDC battery buses 1[2]A and 1[2]B are the site-specific AND                                                                      vital DC buses credited in this EAL.
Indicated voltage is < 112 VDC
: b. Indicated voltage is less than        on both vital 125 VDC buses (site-specific bus voltage value)        1[2]A and 1[2]B for  15 min.
on ALL (site-specific Vital DC           (Note 1) busses) for 15 minutes or longer.
Note   The Emergency Director should       N/A  Note 1: The EC should declare      The classification timeliness note has been standardized across the declare the Unusual Event                          the event promptly       PSL EAL scheme by referencing the "time limit" specified within the promptly upon determining that 15                  upon determining that    EAL wording.
minutes has been exceeded, or                      the time limit has been will likely be exceeded.                           exceeded, or will likely be exceeded.
124 of 124

PSL1                                               GENERAL EMERGENCY                                                                    SITE AREA EMERGENCY                                                                                    ALERT                                                       UNUSUAL EVENT                                                UNIT 1                          GENERAL EMERGENCY                                                        SITE AREA EMERGENCY                                                                              ALERT                                                        UNUSUAL EVENT RG1 Release of gaseous radioactivity resulting in offsite dose greater                   RS1 Release of gaseous radioactivity resulting in offsite dose greater          RA1 Release of gaseous or liquid radioactivity resulting in offsite dose   RU1 Release of gaseous or liquid radioactivity greater than 2 times the                                                                                                                                                                                SA1 Loss of all but one AC power source to emergency buses for           SU1 Loss of all offsite AC power capability to emergency buses for 15 SG1 Prolonged loss of all offsite and all onsite AC power to              SS1 Loss of all offsite power and all onsite AC power to emergency than 1,000 mrem TEDE or 5,000 mrem thyroid CDE                                           than 100 mrem TEDE or 500 mrem thyroid CDE                                      greater than 10 mrem TEDE or 50 mrem thyroid CDE                           ODCM limits for 60 minutes or longer                                                                                                                                                                                                                   15 minutes or longer                                                     minutes or longer emergency buses                                                           buses for 15 minutes or longer RG1.1             1        2        3          4         5      6        DEF            RS1.1         1        2        3           4    5            6      DEF       RA1.1         1       2        3          4     5        6      DEF        RU1.1         1         2     3           4         5       6    DEF SG1.1          1          2      3          4                             SS1.1         1       2       3      4                                  SA1.1         1       2       3          4                               SU1.1              1       2        3          4 Reading on any Table 1R-1 effluent radiation monitor
                                                    > column "GE" for  15 min. (Notes 1, 2, 3, 4)
Reading on any Table 1R-1 effluent radiation monitor
                                                                                                                                            > column "SAE" for  15 min. (Notes 1, 2, 3, 4)
Reading on any Table 1R-1 effluent radiation monitor
                                                                                                                                                                                                                           > column "ALERT" for  15 min. (Notes 1, 2, 3, 4)
Reading on any Table 1R-1 effluent radiation monitor
                                                                                                                                                                                                                                                                                                      > column "UE" for  60 min. (Notes 1, 2, 3)                                             1           Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1A3 and 1B3 Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1A3 and 1B3 for  15 min. (Note 1)
AC power capability, Table 1S-1, to 4.16 kV emergency buses 1A3 and 1B3 reduced to a single power source for Loss of all offsite AC power capability, Table 1S-1, to 4.16 kV emergency buses 1A3 and 1B3 for  15 min. (Note 1)
Loss of        AND                                                                                                                                               15 min. (Note 1)
RG1.2             1        2        3          4         5      6        DEF            RS1.2         1        2        3           4    5            6      DEF       RA1.2         1       2        3          4     5        6      DEF        RU1.2         1        2      3           4         5       6    DEF                 Emergency      REP CETs > 1200&#xba;F                                                                                                                                   AND Reading on any effluent radiation monitors > 2 x the alarm                           AC Power                                                                                                                                                         Any additional single power source failure will result in loss Dose assessment using actual meteorology indicates                                      Dose assessment using actual meteorology indicates                             Dose assessment using actual meteorology indicates                                                                                                                                                                                                                                                                                                                                                                        Table 1S-1 Unit 1 AC Power Sources doses > 1,000 mrem TEDE or 5,000 mrem thyroid                                           doses > 100 mrem TEDE or 500 mrem thyroid CDE at or                            doses > 10 mrem TEDE or 50 mrem thyroid CDE at or                          setpoint established by a current radioactivity discharge                                                                                                                                                                                             of all AC power to SAFETY SYSTEMS CDE at or beyond the SITE BOUNDARY (Note 4)                                             beyond the SITE BOUNDARY (Note 4)                                              beyond the SITE BOUNDARY (Note 4)                                          permit for  60 min.
(Notes 1, 2, 3)                                                                                                                                                                                                                                                                                                                               Offsite:
1                     RG1.3             1        2        3          4         5      6        DEF            RS1.3         1        2        3           4    5            6      DEF       RA1.3         1       2        3          4     5        6      DEF RU1.3         1         2     3           4         5       6    DEF                                SG2 Loss of all emergency AC and vital DC power sources for 15            SS2 Loss of all vital DC power for 15 minutes or longer 1A Startup Transformer 1B Startup Transformer Rad                     Field survey results indicate EITHER of the following at or                             Field survey results indicate EITHER of the following at or                    Analysis of a liquid effluent sample indicates a                                                                                                                                   minutes or longer Effluent                  beyond the SITE BOUNDARY:                                                               beyond the SITE BOUNDARY:                                                      concentration or release rate that would result in doses                   Sample analysis for a gaseous or liquid release indicates                                                                                                                                                                                                                                                                                     Onsite:
Closed window dose rates > 1,000 mR/hr expected to                                       Closed window dose rates > 100 mR/hr expected to                          > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond                         a concentration or release rate 2 x ODCM limits for                                                  SG2.1        1       2       3      4                                    SS2.1         1      2      3     4                                                                                                                               1A Auxiliary Transformer continue for  60 min.
Analyses of field survey samples indicate thyroid CDE continue for  60 min.
Analyses of field survey samples indicate thyroid CDE the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) 60 min. (Notes 1, 2) 2           Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1A3 and 1B3 for  15 min. (Note 1)
Indicated voltage is < 112 VDC on both vital 125 VDC buses 1A and 1B for  15 min. (Note 1) 1B Auxiliary Transformer 1A Diesel Generator 1B Diesel Generator
                                                        > 5,000 mrem for 60 min. of inhalation                                                   > 500 mrem for 60 min. of inhalation                                                                                                                                                                                                       Loss of         AND                                                                                                                                                                               None                                                                    None (Notes 1, 2)                                                                            (Notes 1, 2)                                                                   RA1.4         1       2        3          4     5        6      DEF                                                                                               Vital        Indicated voltage is < 112 VDC on both vital 125 VDC buses                                                                                                                                                                                  1AB - 2AB SBO Cross Tie DC Power       1A and 1B for  15 min. (Note 1)                                                                                                                                                                                                            2A4 (2B4) Cross Tie (if already aligned)
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY (Notes 1, 2):
Closed window dose rates > 10 mR/hr expected to continue  60 min.
R                                                                                                                                                                                                                         Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation Table S-2        Safety System Parameters SA3 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress SU3 UNPLANNED loss of Control Room indications for 15 minutes or longer Abnormal Rad Levels
     /
RG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RS2 Spent fuel pool level at the top of the fuel racks                          RA2 Significant lowering of water level above, or damage to, irradiated fuel RU2 UNPLANNED loss of water level above irradiated fuel 3                                            None Reactor power None SA3.1         1          2    3          4 An UNPLANNED event results in the inability to monitor one SU3.1         1        2       3           4 An UNPLANNED event results in the inability to monitor Rad                                               RG2.1            1       2         3          4        5       6        DEF            RS2.1         1        2        3           4    5            6      DEF      RA2.1          1       2        3          4     5        6      DEF       RU2.1          1        2      3           4         5       6    DEF                 Loss of CR                                                                                                   RCS level                                           or more Table S-2 parameters from within the Control Room                    one or more Table S-2 parameters from within the Control Effluent                                                                                                                                                                                                                                                                                                                                                                                  Indications                                                                                                                                                      for  15 min. (Note 1)                                                       Room for  15 min. (Note 1)
Spent fuel pool level cannot be restored to at least 37 ft.                            Lowering of spent fuel pool level to 37 ft. 3 in. (Level 3) on                IMMINENT uncovery of irradiated fuel in the REFUELING                      UNPLANNED water level drop in the REFUELING PATHWAY                                                                                                                                                RCS pressure                                          AND 3 in. (Level 3) on LI-04-1A or LI-04-1B for  60 min. (Note 1)                         LI-04-1A or LI-04-1B                                                          PATHWAY                                                                    as indicated by any of the following:
Any significant transient is in progress, Table S-3 SFP or reactor cavity level instrument                                                                                                                                                        CET temperature Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds                                                                               RA2.2         1    2       3       4       5      6              DEF            SFP or reactor cavity low level alarm SU4 Reactor coolant activity greater than Technical Specification Damage to irradiated fuel resulting in a release of                             Report of dropping level in SFP or reactor cavity                                                                                                                                             Level in at least one SG allowable limits Release Point & Monitor 2
GE                     SAE                  Alert                      UE                                    radioactivity                                                                AND AND                                                                      UNPLANNED rise in corresponding area radiation levels as                                                                                                                                           Auxiliary feedwater flow to at least                                                                                             SU4.1         1        2        3          4 Plant Vent VALID HIGH alarm on any of the following:                                  indicated by any Table 1R-2 radiation monitors                                                                                                                                                     one SG RCS 26-1 (L/H)                     1.55E+02 &#xb5;Ci/cc      1.55E+01 &#xb5;Ci/cc        1.55E+00 &#xb5;Ci/cc       2.92E-03 &#xb5;Ci/cc                                                                                                                                                                                                                                                                                                                                                                                                                                                 Sample analysis indicates that a reactor coolant activity Irradiated                                                                                                                                                                                            Table 1R-2 area radiation monitors 4
Gaseous ECCS Vent                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                               value is > any of the following allowable limits:
Fuel                                                                                                                                                                                               RSC 26-4 Fuel Handling Building Exhaust Vent                                                                                                                                                                  None                                                                        None                                                                  None RCS 26-2 (L/H) (A)                  7.84E+01 &#xb5;Ci/cc      7.84E+00 &#xb5;Ci/cc        7.84E-01 &#xb5;Ci/cc                N/A                                                                                                                   Table 1R-2 Unit 1 Refueling Area Radiation Monitors                                                                                                                                                                                                                                                                                  60 Ci/gm dose equivalent I-131 instantaneous Events                                   RCS 26-3 (L/H) (B)
RCS                                                                                                                                                                                                                                             1.0 Ci/gm dose equivalent I-131 for > 48 hours FHB Vent                                                                                                                                                 RA2.3          1       2        3          4     5        6      DEF            ARM #          Ch. #                      Description                             Activity                                                                                              Table S-3        Significant Transients                                                                                                    518.9 Ci/gm dose equivalent Xe-133 for > 48 hours RCS 26-4 (L/H)                      2.23E+03 &#xb5;Ci/cc      2.23E+02 &#xb5;Ci/cc        2.23E+01 &#xb5;Ci/cc       9.65E-03 &#xb5;Ci/cc Lowering of spent fuel pool level to 46 ft. 3 in. (Level 2) on                 RE26-4-1        4        Containment CIS Monitor Liquid Waste Discharge                                                                                                                                   LI-04-1A or LI-04-1B                                                                                                                                                                                                                                                         Electrical load rejection > 25% full RE26-6-1        6        Containment CIS Monitor RSC-26-43                                N/A                    N/A                   N/A            1.33E-03 &#xb5;Ci/ml                                                                                                                                                                                                                                                                                                               electrical load Liquid RE26-7-1        7        Spent Fuel Pool SG Blowdown                                                                                                                                                                                                                             RE26-8-1        8        Refueling Canal                                                                                                                                                      Reactor Trip                                                                                                                  SU5 RCS leakage for 15 minutes or longer A SG RE-26-44-1A                        N/A                    N/A                   N/A            2.44E+04 cpm B SG RE-26-45-1B                        N/A                    N/A                   N/A            5.81E+04 cpm                                                                                                                    RE26-36-1       36       Operating Deck SIAS Activation                                                                                                               SU5.1        1     2      3      4 5
RCS unidentified or pressure boundary leakage RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                     > 10 gpm for  15 min.
None                                                                     None                                                                  None                                    OR Table R-3 Safe Operation & Shutdown Rooms Areas                                                                                                                                                                                                                                                                                         RCS                                                                                                                                                                                                                                    RCS identified leakage > 25 gpm for  15 min.
RA3.1          1       2        3          4     5        6      DEF 3
Leakage                                                                                                                                                                                                                                   OR Room/Area                                    Mode                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Leakage from the RCS to a location outside containment Dose rates > 15 mR/hr in EITHER of the following:
S Reactor Auxiliary Building (0.5None and 19.5 levels)          1, 2, 3, 4, 5                                               None                                           Control Room                                                                                              None                                                                                                                                                                                                                                                                                        > 25 gpm for  15 min.
Area                                                                                                                                                                                                 Central Alarm Station                                                                                                                                                                                                                                                                                                                                                                                 (Note 1)
Turbine Building (east end levels required for Radiation                         access to AFW, Steam Trestle, Switchgear and              1, 2, 3, 4, 5 Levels                                                                                                                                                                                           RA3.2          1       2        3          4     5                                                                                                   System                                                                                               SS6 Inability to shut down the reactor causing a challenge to core      SA6 Automatic or manual trip fails to shut down the reactor and          SU6 Automatic or manual trip fails to shut down the reactor Control Rooms) cooling or RCS heat removal                                             subsequent manual actions taken at the reactor control consoles Cable Spreading Room                                      1, 2, 3, 4, 5                                                                                         An UNPLANNED event results in radiation levels that                                                                                                  Malfunct.
are not successful in shutting down the reactor prohibit or IMPEDE access to any Table R-3 room or area A3/B3 Switchgear Rooms                                    1, 2, 3, 4, 5 (Note 5)                                                                                                                                                                                                                                                  SS6.1          1        2                                               SA6.1         1          2                                                   SU6.1         1        2 Steam Trestle Area                                        1, 2, 3, 4, 5 Intake Structure (Note 15)                                1, 2, 3, 4, 5         HS1 HOSTILE ACTION within the PROTECTED AREA                                    HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or                     HU1 Confirmed SECURITY CONDITION or threat                                                                                                                                     An automatic or manual trip did not shut down the reactor as            An automatic or manual trip (RX Trip Pushbuttons) did not                An automatic trip did not shut down the reactor as airborne attack threat within 30 minutes                                                                                                                                                                                                              indicated by reactivity control Safety Function Status Check            shut down the reactor as indicated by reactivity control                 indicated by reactivity control Safety Function Status Component Cooling Water Area                              1, 2, 3, 4, 5 acceptance criteria not met                                             Safety Function Status Check acceptance criteria not met                 Check acceptance criteria not met after any RPS Emergency Diesel Generator Buildings                      1, 2, 3, 4, 5         HS1.1           1        2        3           4    5            6      DEF      HA1.1          1       2        3          4     5        6      DEF       HU1.1          1         2      3          4         5       6    DEF AND                                                                     AND                                                                    setpoint is exceeded 1                                                                                                                                                                                                                                                                                                                                                               6 Refueling Water Tank                                      1, 2, 3, 4, 5         A HOSTILE ACTION is occurring or has occurred within the                        A HOSTILE ACTION is occurring or has occurred within                       A SECURITY CONDITION that does not involve a HOSTILE                                                                                      None                                 All actions taken to shut down the reactor are not successful           Subsequent automatic or manual trip actions (RX Trip                       AND Water Treatment Plant/Storage Tanks                        1, 2, 3, 4           PROTECTED AREA as reported by PSL Security Shift                                the OWNER CONTROLLED AREA as reported by PSL                               ACTION as reported by PSL Security Shift Supervision                                                                                                                           as indicated by reactivity control Safety Function Status               Pushbuttons) are not successful in shutting down the                     A subsequent automatic trip or manual trip (RX Trip None                                             Supervision                                                                     Security Shift Supervision                                                  OR                                                                                                                                                                            Check acceptance criteria not met                                       reactor as indicated by reactivity control Safety Function               Pushbuttons) are successful in shutting down the reactor Security                         Containment                                                     4, 5 OR                                                                       Notification of a credible security threat directed at the site                          RPS                                                                                     AND EITHER:                                                           Status Check acceptance criteria not met (Note 8)                        as indicated by reactivity control Safety Function Status A validated notification from NRC of an aircraft attack                     OR                                                                                     Failure                                                                                                                                                                                                                                 Check acceptance criteria met (Note 8)
REP CETs >1200&deg;F threat within 30 min. of the site                                          A validated notification from the NRC providing information of                                                                                                                                                                                                                                                                   SU6.2        1      2 NOTES                                                                                                                                                                     an aircraft threat                                                                                                                                                                  Applicable RCS and Core Heat Removal (HR) Safety A manual trip (RX Trip Pushbuttons) did not shut down the Function Status Check acceptance criteria not met Note 1:      The EC should declare the event promptly upon determining that the time                                                                                                                                                                                                                                                                                                                                                                                                                                                                           reactor as indicated by reactivity control Safety Function limit has been exceeded, or will likely be exceeded.                                                                                         [Refer to CA6.1 or SA9.1 for potential escalation                HU2 Seismic event greater than OBE levels                                                                                                                                                                                                                                                                                           Status Check acceptance criteria not met (Note 8) 2                                           Note 2:      If an ongoing release is detected and the release start time is unknown, assume None  that the release duration has exceeded the specified time None limit.
due to a seismic event]
None HU2.1          1        2      3           4         5       6    DEF AND A subsequent automatic trip or manual trip action (RX Trip Pushbuttons) is successful in shutting down the reactor as Seismic                                                                                                                                                                                                                                                                    Seismic event > OBE as indicated by S-36 CNTMT SEISMIC                                                                                                                                                                                                                                                                              indicated by reactivity control Safety Function Status Check Event                                          Note 3:      If the effluent flow past an effluent monitor is known to have stopped due to                                                                                                                                 MONITOR HIGH LIMIT alarm                                                                                                                                                                                                                                                                                                            acceptance criteria met actions to isolate the release path, then the effluent monitor reading is no                                                                                                                                                                                                                                                                                                                                                                                         Table S-4      Communications Methods longer VALID for classification purposes.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         SU7 Loss of all onsite or offsite communications capabilities HU3 Hazardous event
[Refer to CA6.1 or SA9.1 for potential escalation                                                                                                                                                                                                                                                                               System                     Onsite    Offsite      NRC Note 4:      The pre-calculated effluent monitor values presented in EALs RA1.1, 7
due to a hazardous event]                                                                                                                                                                                                                                                                                                                                                                 SU7.1        1      2      3      4 RS1.1 and RG1.1 should be used for emergency classification                                                                                                                                                                                                                                                                                                                                                                                            Plant Radio System                                     X HU3.1          1         2     3           4         5       6    DEF                                                                                                                                                                                                                                                               Loss of all Table S-4 onsite communication methods assessments until the results from a dose assessment using actual                                                                                                                                                                                                                                                                                       None                                                                     None                                                                  None meteorology are available.                                                                                                                                                                                    A tornado strike within the PROTECTED AREA                                                                                                                                                                                               Commercial Telephone System                            X       X          X                 OR Loss of Plant Address (PA) System                              X                                   Loss of all Table S-4 State and local agency communication Commun.
Note 5:      If the equipment in the listed room or area was already inoperable or out-                                                                                                                                    HU3.2          1        2      3           4         5       6    DEF                                                                                                                                                                                                                                                               methods Federal Telephone System (ENS)                                            X                 OR of-service before the event occurred, then no emergency classification is                                                                                                                                     Internal room or area FLOODING of a magnitude sufficient to warranted.                                                                                                                                                                                                                                                                                                                                                                                                                                             EMnet                                                          X                           Loss of all Table S-4 NRC communication methods require manual or automatic electrical isolation of a SAFETY 3                                           Note 6:      If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute None time limit, declaration of a General Emergency is not required None                                                                                               None SYSTEM component required by technical specifications for the current operating mode Hot Ring Down (HRD) Telephone System Shift Manager Cell Phones X
X          X SU8 Failure to isolate containment or loss of containment pressure control Natural or                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    SU8.1             1       2        3           4 8
Tech.                                            Note 7:      This EAL does not apply to routine traffic impediments such as fog, snow,                                                                                                                                     HU3.3          1        2      3           4         5       6    DEF Hazard                                                         ice, or vehicle breakdowns or accidents.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      Any penetration is not closed within 15 min. of a VALID Movement of personnel within the PROTECTED AREA is None                                                                     None                                                                  None                                   CIS actuation signal Note 8:      A manual trip action is any operator action, or set of actions, which causes                                                                                                                                  IMPEDED due to an event external to the PROTECTED CTMT                                                                                                                                                                                                                                      OR the control rods to be rapidly inserted into the core, and does not include                                                                                                                                   AREA involving hazardous materials (e.g., an offsite chemical Failure                                                                                                                                                                                                                                 CTMT pressure > 10 psig with < one full train of CTMT manually driving in control rods or implementation of boron injection                                                                                                                                         spill or toxic gas release) heat removal systems operating per design for  15 min.
strategies.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                   (Notes 1, 11)
HU3.4          1        2      3           4         5       6    DEF Note 9:      If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency                                                                                                                                                   A hazardous event that results in on-site conditions sufficient classification is not warranted.                                                                                                                                                                              to prohibit the plant staff from accessing the site via personal                                                                                                                                                                                       SA9 Hazardous event affecting SAFETY SYSTEMS needed for the vehicles (Note 7)                                                                                                                                                                                                                                          current operating mode Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no                                                                                                                                                                                                                                                                                                                                                     Table S-5    Hazardous Events indications of degraded performance to at least one train of a SAFETY                                                                                                                                             HU4 FIRE potentially degrading the level of safety of the plant                                                                                                                                                                                         SA9.1        1     2        3    4 SYSTEM, then this emergency classification is not warranted.                                                                               [Refer to CA6.1 or SA9.1 for potential escalation due to fire]
Seismic event (earthquake)                                    The occurrence of any Table S-5 hazardous event Note 11: One full train of containment heat removal systems consist of any one of the following:
HU4.1          1         2     3           4         5 A FIRE is not extinguished within 15 min. of any of the 6    DEF 9                                                                                               Internal or external FLOODING event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed Two (2) Containment Spray Headers with flow in each header at least                                                                                                                                      following fire detection indications (Note 1):                                        Hazardous                                           None                                                                                                                                                                                                                          None Table H-1 PSL Fire Areas                                    Report from the field (i.e., visual observation)                                                                                                                                     High winds or tornado strike                                  for the current operating mode 2,700 gpm                                                                                                                                                                                                                                                                                        Event Receipt of multiple (more than 1) fire zone alarms or                                                                                                                                                                                                AND EITHER:
Four (4) Containment Coolers                                                                                                                                                                                                                                                                    Affecting                                                                                          FIRE Reactor Containment Building/Shield Building                     indications                                                                                                                                                                                                                                             Event damage has caused indications of degraded Safety Two (2) Containment Coolers with at least one (1) Containment Spray                                                                          Control Room                                                     Field verification of a single fire zone alarm                                                                                                                                                                                                          performance to the second train of the SAFETY Systems                                                                                            EXPLOSION Header with at least 2,700 gpm                                                                                                               Reactor Auxiliary Building (RAB)                             AND                                                                                                                                                                                                                                                         SYSTEM needed for the current operating mode Diesel Generator Building                                                                                                                                                                                                                                                                                                Event damage has resulted in VISIBLE DAMAGE to Note 12: This threshold is only applicable if all RVLMS Sensors 4 through 8 are not                                                                                                                                        The FIRE is located in any Table H-1 area                                                                                                                                                 Other events with similar hazard characteristics Component Cooling Water (CCW)                                                                                                                                                                                                                                                                                            the second train of the SAFETY SYSTEM needed for covered inclusive.                                                                                                                                    Intake Structure                                                                                                                                                                                                                                      as determined by the Shift Manager Intake Cooling Water (ICW)
HU4.2          1         2      3          4         5       6    DEF                                                                                                                                                                                        the current operating mode Note 13: If an RCS heat removal system is in operation within the applicable Table                                                                                                                                                                                                                                                                                                                                                                                                 (Notes 9, 10) 4 Steam Trestle Area                                          Receipt of a single fire zone alarm (i.e., no other indications of H                                                                            C-4  heat-up duration and RCS temperature is being reduced, the EAL None                                                           None is                                                                             Yard Areas a FIRE)
F not applicable.                                                                                                                                       o RWT AND                                                                                                 FG1.1         1          2      3          4                              FS1.1         1        2       3          4                             FA1.1        1          2    3          4 Fire                                                                                                                                                                                                       o CST Hazards                                                                 Note 14: A containment fire zone alarm is considered VALID upon receipt of multiple                                                                            o DOST                                                      The fire zone alarm is indicating a FIRE within any Table H-1                                         Loss of any two barriers                                                  Loss or potential loss of any two barriers (Table F-1)                  Any loss or any potential loss of EITHER Fuel Clad or                                                           None zones (more than one) actuated on CP-376 panel.                                                                                                                                                                   area (excluding containment) (Note 14)
Fission Product              AND                                                                                                                                             RCS (Table F-1)
Note 15: The Intake Structure includes the Intake Cooling Waiter (ICW) system and                                                                                                                                            AND                                                                                                 Loss or potential loss of the third barrier (Table F-1)
Barrier Degradation those immediately surrounding areas that could potentially affect                                                                                                                                                 The existence of a FIRE is not verified (proved or disproved) operability or access to ICW piping and components needed for safe                                                                                                                                                within 30 min. of alarm receipt (Note 1) shutdown.                                                                                                                                                                                                         HU4.3          1         2     3           4         5       6    DEF A FIRE within the PROTECTED AREA not extinguished within Table H-2      Safe Operation & Shutdown Rooms/Areas                                                                                                                60 min. of the initial report, alarm or indication (Note 1)
Room/Area                                  Mode HU4.4          1         2     3           4         5       6    DEF Table F-1 Fission Product Barrier Matrix A FIRE within the PROTECTED AREA that requires Reactor Auxiliary Building (0.5 and 19.5 levels)            1, 2, 3, 4, 5                                                                                           firefighting support by an offsite fire response agency to Turbine Building (east end levels required for access to    1, 2, 3, 4, 5                                                                                           extinguish                                                                                                                         Fuel Clad (FC) Barrier                                                           Reactor Coolant System (RCS) Barrier                                                                     Containment (CTMT) Barrier AFW, Steam Trestle, Switchgear and Control Rooms)
HA5 Gaseous release IMPEDING access to equipment necessary for Cable Spreading Room                                        1, 2, 3, 4, 5 normal plant operations, cooldown or shutdown                                                                                                       Category                                Loss                                     Potential Loss                                              Loss                               Potential Loss                                              Loss                                           Potential Loss A3/B3 Switchgear Rooms                                      1, 2, 3, 4, 5 Steam Trestle Area                                          1, 2, 3, 4, 5                HA5.1          1        2       3          4      5                                                                                                                                                                                                                              1. An automatic or manual SIAS                 1. UNISOLABLE RCS or SG tube                      1. A leaking or RUPTURED SG is 5
: 1. RVLMS indicates Sensors 4 through 8 Intake Structure (Note 15)                                  1, 2, 3, 4, 5                Release of a toxic, corrosive, asphyxiant or flammable gas                                                                                                                                                                                                                          actuation required by EITHER:                  leakage > 50 gpm                                  FAULTED outside of CTMT not covered (Note 12) into any Table H-2 room or area                                                                                                                                                                                                                                                      UNISOLABLE RCS leakage Component Cooling Water Area                                1, 2, 3, 4, 5 A
None                                                                                None                                                                                                                                                        None SG tube RUPTURE                            2. Uncontrolled RCS cooldown Hazardous                                                                                                 Emergency Diesel Generator Buildings                        1, 2, 3, 4, 5 AND None                                                                                                                            > 100&#xba;F/hr and RCS pressure and                                                                                            None Gases                                                                                                                                                                                            Entry into the room or area is prohibited or IMPEDED                                                                                                   RCS or SG Tube Refueling Water Tank                                        1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                   temperature are to the left of the (Note 5)                                                                                                                                                  Leakage Water Treatment Plant/Storage Tanks                          1, 2, 3, 4 Maximum Subcooled (PT) Curve (EOP Figure 1A or 1B)
Containment                                                       4, 5 HS6 Inability to control a key safety function from outside the Control         HA6 Control Room evacuation resulting in transfer of plant control to                                                                                                          1. REP CETs > 1200&deg;F                             2. REP CETs > 700&deg;F                                                                            3. Both SG wide range level < 15% for                                                                      1. REP CETs > 1200&deg;F Room                                                                            alternate locations                                                                                                                                                                                                                                                                                                            15 min.                                                                                                    AND
: 3. Both SG wide range level < 15% for                                                              AND                                                                                                       Functional Recovery (1-EOP-15)
HS6.1            1       2        3            4    5           6               HA6.1          1       2        3          4     5        6      DEF                                                                                         B                                                                      15 min.                                                             None OTC not established None for RCS and Core Heat Removal Inadequate Heat                                                              AND                                                                                                                                                                                                      not effective within 15 min.
6 An event has resulted in the Control Room being evacuated                       An event has resulted in the Control Room being evacuated Removal                                                                 OTC not established                                                                                                                                                                                       (Note 1)
None AND                                                                                                                                                                                          None Control of any of the following key safety functions is not Control re-established within 15 min. of the last licensed operator Room leaving the Control Room (Note 1):
Evacuation                                                                                                                                                                                                                                                                                                                                                                        2. CTMT high range radiation                                                                      2. Any CIAS monitor Ch. # 3/4/5/6                                                                                                                        2. CTMT high range radiation Reactivity (modes 1, 2 and 3 only) monitors Ch. # 58/59                                                                              > 1.5E+03 mR/hr                                                                                                                                          monitors Ch. # 58/59 Core cooling
                                                                                                                                                                                                                                                                                                                                                                                                            > 1.4E+02 R/hr                                                                                                                                                                                                                                             > 2.7E+03 R/hr RCS heat removal C                 3. Coolant activity > 300 Ci/gm dose equivalent I-131                                                 None                                                                                             None                                                None CTMT Radiation /
HG7 Other conditions exist which in the judgment of the EC warrant                       HS7 Other conditions existing that in the judgment of the EC warrant            HA7 Other conditions exist that in the judgment of the EC warrant          HU7 Other conditions existing that in the judgment of the EC warrant          RCS Activity declaration of a General Emergency                                                       declaration of a Site Area Emergency                                            declaration of an Alert                                                    declaration of an Unusual Event HG7.1              1       2         3          4         5      6        DEF          HS7.1           1        2        3            4    5           6      DEF      HA7.1          1       2        3          4     5        6      DEF       HU7.1          1         2      3          4         5       6    DEF Other conditions exist which, in the judgment of the EC,                               Other conditions exist which, in the judgment of the EC,                        Other conditions exist which, in the judgment of the EC,                   Other conditions exist which, in the judgment of the EC, 7                    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 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 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 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
: 2. CTMT isolation (CIS) is required AND EITHER:
: 3. CTMT pressure > 42 psig EC                                                                                                             results in intentional damage or malicious acts, (1) toward site HOSTILE ACTION that results in an actual loss of physical                                                                                                              probable life threatening risk to site personnel or damage to              been initiated. No releases of radioactive material requiring                                                                                                                                                                                                                                                  CTMT integrity has been lost                    4. CTMT hydrogen concentration Judgment control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.                                             D                                                                                                                                                                                                                            based on EC judgment                               > 4%
CTMT Integrity or                           None                                            None                                                  None                                       None                                 UNISOLABLE pathway from                          5. CTMT pressure > 10 psig with offsite for more than the immediate site area.                                         the protection of the public. Any releases are not expected                     EPA Protective Action Guideline exposure levels.
Bypass                                                                                                                                                                                                                         CTMT atmosphere to the                              < one full train of CTMT heat to result in exposure levels which exceed EPA Protective environment exists                                  removal systems operating per Action Guideline exposure levels beyond the SITE BOUNDARY.                                                                                                                                                                                                                                                                                                                                                                                                                                                         3. Indications of RCS leakage outside                      design for  15 min.
of CTMT                                                 (Notes 1, 11)
EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EU1.1        1      2      3       4     5      6     DEF                                           4. Any condition in the opinion of the           4. Any condition in the opinion of the EC        3. Any condition in the opinion of the         4. Any condition in the opinion of the            4. Any condition in the opinion of the                  6. Any condition in the opinion of the E
E                       1                                                             None                                                                                None                                                                          None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HSM) > any of the following:
EC Judgment EC that indicates loss of the Fuel Clad barrier that indicates potential loss of the Fuel Clad barrier EC that indicates loss of the RCS barrier EC that indicates potential loss of the RCS barrier EC that indicates loss of the CTMT barrier EC that indicates potential loss of the CTMT barrier Confinement ISFSI              Boundary                                                                                                                                                                                                                                                                          1,600 mrem/hr on the HSM front bird screen 10 mrem/hr on the outside HSM door centerline 10 mrem/hr on the HSM end shield wall exterior St. Lucie Plant Modes:                                                        1 Power Operation 2
Startup 3
Hot Standby 4
Hot Shutdown             Cold Shutdown 5                        6 Refueling DEF Defueled UNIT 1 Emergency Action Level Matrix                                                                 EAL WALL CHART - HOT CONDITIONS (RCS > 200&deg;F)
Revision X Prepared for FPL by Operations Support Services, Inc. 4-19-19

PSL1                                                GENERAL EMERGENCY                                                                 SITE AREA EMERGENCY                                                                                     ALERT                                                       UNUSUAL EVENT                                                UNIT 1                        GENERAL EMERGENCY                                                         SITE AREA EMERGENCY                                                                        ALERT                                                      UNUSUAL EVENT RG1 Release of gaseous radioactivity resulting in offsite dose greater                 RS1 Release of gaseous radioactivity resulting in offsite dose greater          RA1 Release of gaseous or liquid radioactivity resulting in offsite dose   RU1 Release of gaseous or liquid radioactivity greater than 2 times the                            CG1 Loss of RCS inventory affecting fuel clad integrity with               CS1 Loss of RCS inventory affecting core decay heat removal         CA1 Significant loss of RCS inventory                                    CU1 UNPLANNED loss of RCS inventory than 1,000 mrem TEDE or 5,000 mrem thyroid CDE                                         than 100 mrem TEDE or 500 mrem thyroid CDE                                      greater than 10 mrem TEDE or 50 mrem thyroid CDE                           ODCM limits for 60 minutes or longer                                                               containment challenged                                                     capability RG1.1              1       2         3         4     5       6           DEF         RS1.1          1       2        3            4    5          6        DEF       RA1.1         1       2        3           4    5        6      DEF        RU1.1         1         2     3           4         5       6    DEF CG1.1                                             5       6                CS1.1                                        5       6              CA1.1                                              5    6                CU1.1                                           5        6 Reading on any Table 1R-1 effluent radiation monitor                                 Reading on any Table 1R-1 effluent radiation monitor                            Reading on any Table 1R-1 effluent radiation monitor                       Reading on any Table 1R-1 effluent radiation monitor
                                                    > column "GE" for  15 min. (Notes 1, 2, 3, 4)                                       > column "SAE" for  15 min. (Notes 1, 2, 3, 4)                                 > column "ALERT" for  15 min. (Notes 1, 2, 3, 4)                          > column "UE" for  60 min. (Notes 1, 2, 3)                                                       RCS level cannot be monitored for  30 min. (Note 1)                       0 in. on LI-1117 or LI-1117-1                                       Loss of RCS inventory as indicated by RCS water level                    UNPLANNED loss of reactor coolant results in RCS water AND                                                                                                                                           < minimum required for continued LPSI pump operation                     level less than a required lower limit for  15 min. (Note 1) 1        2      3           4         5       6    DEF                              Core uncovery is indicated by any of the following:                        CS1.2                                        5       6 RG1.2              1       2         3         4     5       6           DEF         RS1.2          1       2        3            4    5          6        DEF       RA1.2         1       2        3           4    5        6      DEF        RU1.2 UNPLANNED increase in any Table C-1 sump or tank                                                                                            CA1.2                                              5    6                CU1.2                                           5        6 RCS level cannot be monitored for  30 min. (Note 1) 1 Dose assessment using actual meteorology indicates                                   Dose assessment using actual meteorology indicates                              Dose assessment using actual meteorology indicates                         Reading on any effluent radiation monitors > 2 x the alarm                                           level of sufficient magnitude to indicate core uncovery doses > 1,000 mrem TEDE or 5,000 mrem thyroid                                        doses > 100 mrem TEDE or 500 mrem thyroid CDE at or                             doses > 10 mrem TEDE or 50 mrem thyroid CDE at or                          setpoint established by a current radioactivity discharge                                                                                                                     AND Visual observation of UNISOLABLE RCS leakage of                                                                                             RCS water level cannot be monitored for  15 min. (Note 1)               RCS level cannot be monitored CDE at or beyond the SITE BOUNDARY (Note 4)                                          beyond the SITE BOUNDARY (Note 4)                                               beyond the SITE BOUNDARY (Note 4)                                          permit for  60 min.                                                                                                                                                         Core uncovery is indicated by any of the following:
sufficient magnitude to indicate core uncovery                                                                                               AND EITHER                                                               AND EITHER:
(Notes 1, 2, 3)                                                                      RCS Level                                                                                  UNPLANNED increase in any Table C-1 sump or tank 1                     RG1.3              1       2         3         4     5       6           DEF         RS1.3          1       2        3            4    5          6        DEF       RA1.3         1       2        3           4    5        6      DEF RU1.3         1         2     3           4         5       6    DEF Containment High Range Radiation Monitor reading
                                                                                                                                                                                                                                                                                                                                                                                                         > 4.8E+04 R/hr (Refueling Mode)
Erratic source range monitor indications level of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Rad                     Field survey results indicate EITHER of the following at or                          Field survey results indicate EITHER of the following at or                     Analysis of a liquid effluent sample indicates a                                                                                                                                                                                                           sufficient magnitude to indicate core uncovery Sample analysis for a gaseous or liquid release indicates                                          AND Effluent                  beyond the SITE BOUNDARY:                                                            beyond the SITE BOUNDARY:                                                       concentration or release rate that would result in doses                                                                                                                                                                                                   Containment High Range Radiation Monitor reading a concentration or release rate 2 x ODCM limits for                                               Any Containment Challenge indication, Table C-2 Closed window dose rates > 1,000 mR/hr expected to                                  Closed window dose rates > 100 mR/hr expected to                            > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond                                                                                                                                                                                                         > 4.8E+04 R/hr (Refueling Mode) continue for  60 min.                                                              continue for  60 min.                                                      the SITE BOUNDARY for 60 min. of exposure                                   60 min. (Notes 1, 2)                                                                                                                                                          Erratic source range monitor indications Analyses of field survey samples indicate thyroid CDE                               Analyses of field survey samples indicate thyroid CDE                       (Notes 1, 2)
                                                         > 5,000 mrem for 60 min. of inhalation                                              > 500 mrem for 60 min. of inhalation                                                                                                                                                                                                                           Table C-2      Containment Challenge Indications                                        Table C-1      Sumps/Tanks (Notes 1, 2)                                                                         (Notes 1, 2)                                                                    RA1.4         1       2        3           4    5        6      DEF Field survey results indicate EITHER of the following at or                                                                                                                             CONTAINMENT CLOSURE not established                                                     Containment Sump                            CA2 Loss of all offsite and all onsite AC power to emergency buses for   CU2 Loss of all but one AC power source to emergency buses for 15 beyond the SITE BOUNDARY (Notes 1, 2):                                                                                                                                                  (Note 6)                                                                                                                                15 minutes or longer                                                     minutes or longer Reactor Cavity Sump Closed window dose rates > 10 mR/hr expected to continue  60 min.                                                                                                                                                                  CTMT hydrogen concentration > 4%                                                        Safeguards Sump                             CA2.1                                              5    6        DEF     CU2.1                                           5        6      DEF R                                                                                                                                                                                                                       Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation UNPLANNED increase in CTMT pressure None Holdup Tanks None Refueling Water Tank Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 1A3 and 1B3 for  15 min. (Note 1)
AC power capability, Table 1C-3, to 4.16 kV emergency buses 1A3 and 1B3 reduced to a single power source for Abnormal Rad Levels RG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RS2 Spent fuel pool level at the top of the fuel racks                           RA2 Significant lowering of water level above, or damage to, irradiated fuel RU2 UNPLANNED loss of water level above irradiated fuel                                  2                     Table 1C-3 Unit 1 AC Power Sources CCW Surge Tank Quench Tank 15 min. (Note 1)
AND Loss of                                                                                                                                                                                                                             Any additional single power source failure will result in
     /                                                                                                                                                                                                                                                                                                                                                                                   Emergency Equipment Drain Tank loss of all AC power to SAFETY SYSTEMS Rad                                               RG2.1             1       2        3          4     5       6           DEF         RS2.1          1       2        3            4    5          6        DEF      RA2.1          1       2        3           4    5        6      DEF       RU2.1          1        2      3           4         5       6    DEF                 AC Power                 Offsite:
Effluent                                            Spent fuel pool level cannot be restored to at least 37 ft.                         Lowering of spent fuel pool level to 37 ft. 3 in. (Level 3) on                 IMMINENT uncovery of irradiated fuel in the REFUELING                      UNPLANNED water level drop in the REFUELING PATHWAY                                                               1A Startup Transformer 3 in. (Level 3) on LI-04-1A or LI-04-1B for  60 min. (Note 1)                      LI-04-1A or LI-04-1B                                                           PATHWAY                                                                    as indicated by any of the following:                                                                             1B Startup Transformer SFP or reactor cavity level instrument                                                                       1A Auxiliary Transformer back fed via Table 1R-1 Unit 1 Effluent Monitor Classification Thresholds                                                                             RA2.2         1    2       3       4       5      6              DEF            SFP or reactor cavity low level alarm                                                                        Main Transformer (if already aligned)
Damage to irradiated fuel resulting in a release of                             Report of dropping level in SFP or reactor cavity                                                            1B Auxiliary Transformer back fed via Release Point & Monitor                      GE                      SAE                                                                                                                                                                                                                                                                                 Main Transformer (if already aligned) 2 Alert                      UE                                     radioactivity                                                                AND Plant Vent                                                                                                                                               AND                                                                      UNPLANNED rise in corresponding area radiation levels as                                                                                                                                                                                          CA3 Inability to maintain plant in cold shutdown                         CU3 UNPLANNED increase in RCS temperature RCS 26-1 (L/H)                 1.55E+02 &#xb5;Ci/cc      1.55E+01 &#xb5;Ci/cc         1.55E+00 &#xb5;Ci/cc       2.92E-03 &#xb5;Ci/cc                                   VALID HIGH alarm on any of the following:                                  indicated by any Table 1R-2 radiation monitors                                                                 Onsite:
1A Diesel Generator C
Irradiated                                                                                                                                                                                          Table 1R-2 area radiation monitors Gaseous ECCS Vent                                                                                                                                                                                                                                                                                                                                          1B Diesel Generator                                                                                                              CA3.1                                              5    6                CU3.1                                           5        6 Fuel                                                                                                                                                                                             RSC 26-4 Fuel Handling Building Exhaust Vent 3
RCS 26-2 (L/H) (A)              7.84E+01 &#xb5;Ci/cc      7.84E+00 &#xb5;Ci/cc         7.84E-01 &#xb5;Ci/cc                N/A                                                                                                                    Table 1R-2 Unit 1 Refueling Area Radiation Monitors                                                           1AB - 2AB SBO Cross Tie Events                                   RCS 26-3 (L/H) (B)                                                                                                                                                                                                                                                                                                                                                                                                                                                                UNPLANNED rise in RCS temperature to > 200&deg;F for                         UNPLANNED rise in RCS temperature to > 200&deg;F due to Cold S/D                               2A4 (2B4) Cross Tie (if already aligned)
None                                                                                                               > Table C-4 duration due to a loss of RCS cooling                        loss of decay heat removal capability FHB Vent                                                                                                                                              RA2.3          1       2        3           4    5        6      DEF            ARM #          Ch. #                      Description                / Refuel RCS 26-4 (L/H)                 2.23E+03 &#xb5;Ci/cc      2.23E+02 &#xb5;Ci/cc         2.23E+01 &#xb5;Ci/cc       9.65E-03 &#xb5;Ci/cc                                                                                                                                                                                                       RCS                                                                                                                                                      (Notes 1, 13)
Lowering of spent fuel pool level to 46 ft. 3 in. (Level 2) on                 RE26-4-1        4        Containment CIS Monitor                      System        Temp.                                                                                                                                                       OR                                                                     CU3.2                                           5       6 Liquid Waste Discharge                                                                                                                                 LI-04-1A or LI-04-1B                                                                                                                                 Malfunct.                                                                                                                                                               UNPLANNED RCS pressure rise > 10 psig RE26-6-1        6        Containment CIS Monitor                                                                                                                                                                                                                                                                       Loss of all RCS temperature and RCS water level indication RSC-26-43                            N/A                     N/A                   N/A            1.33E-03 &#xb5;Ci/ml                                                                                                                                                                                                                                                                                                                                                                (does not apply to solid plant conditions)
Liquid RE26-7-1        7        Spent Fuel Pool                                                                                                                                                                                                                                                                               for  15 min. (Note 1)
SG Blowdown                                                                                                                                                                                                                           RE26-8-1        8        Refueling Canal Table C-4 RCS Heatup Duration Thresholds A SG RE-26-44-1A                    N/A                     N/A                   N/A            2.44E+04 cpm B SG RE-26-45-1B                    N/A                     N/A                   N/A            5.81E+04 cpm                                                                                                                     RE26-36-1       36       Operating Deck                                                                                                                                                                                                                                                                                CU4 Loss of vital DC power for 15 minutes or longer CONTAINMENT CLOSURE RCS Status                                               Heat-up Duration Status                                                                                                                                                          CU4.1                                           5        6 RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown                                                                                                                   4                                   Intact AND not reducedNone inventory N/A                        60 minutes       None Indicated voltage is < 112 VDC on required vital 125 VDC buses 1A and 1B for  15 min. (Note 1)
Table R-3 Safe Operation & Shutdown Rooms Areas                                                                                                                                                                                                                                                                                      Loss of RA3.1          1       2        3           4    5        6      DEF                                                                                              Vital DC                                                                                              20 minutes 3
Established Room/Area                               Mode                                                                                                                                                                                                                                                                  Power                                Not intact OR                                                                                                       Table C-5      Communications Methods Dose rates > 15 mR/hr in EITHER of the following:
reduced inventory Reactor Auxiliary Building (0.5None and 19.5 levels)      1, 2, 3, 4, 5                                                None                                            Control Room                                                                                              None Not Established                   0 minutes Area                                                                                                                                                                                               Central Alarm Station Turbine Building (east end levels required for                                                                                                                                                                                                                                                                                                                                                                                                                                                   System                      Onsite      Offsite       NRC Radiation access to AFW, Steam Trestle, Switchgear and          1, 2, 3, 4, 5                                                                                           RA3.2          1       2        3           4    5                                                                                                                                                                                                                                                                                                                                                    CU5 Loss of all onsite or offsite communications capabilities Levels                           Control Rooms)                                                                                                                                                                                                                                                                                                                                                                                                                                                              Plant Radio System                                  X Cable Spreading Room                                  1, 2, 3, 4, 5                                                                                           An UNPLANNED event results in radiation levels that                                                                                                                                                                                                                                                           Commercial Telephone System                         X             X         X           CU5.1                                           5       6       DEF 5
prohibit or IMPEDE access to any Table 1R-3 room or                                                                                                                                                                                                                                                           Plant Address (PA) System                           X A3/B3 Switchgear Rooms                                1, 2, 3, 4, 5 area (Note 5)                                                                                                                                                                                                                                                                                                                                                                                         Loss of all Table C-5 onsite communication methods Steam Trestle Area                                    1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                           None                                                                     None                 Federal Telephone System (ENS)                                              X OR HS1 HOSTILE ACTION within the PROTECTED AREA                                    HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or                     HU1 Confirmed SECURITY CONDITION or threat                                             Loss of                                                                                                                                     EMnet                                                             X Intake Structure (Note 15)                            1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                 Loss of all Table C-5 State and local agency communication airborne attack threat within 30 minutes Component Cooling Water Area                          1, 2, 3, 4, 5                                                                                                                                                                                                                                                            Commun.                                                                                                                                      Hot Ring Down (HRD) Telephone System                              X                     methods HS1.1           1       2        3            4    5          6        DEF      HA1.1          1       2        3           4    5        6      DEF       HU1.1          1         2      3          4         5       6    DEF                                                                                                                                                              Shift Manager Cell Phones                                         X         X            OR Emergency Diesel Generator Buildings                  1, 2, 3, 4, 5 Loss of all Table C-5 NRC communication methods 1                           Refueling Water Tank Water Treatment Plant/Storage Tanks None 1, 2, 3, 4, 5 1, 2, 3, 4 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by PSL Security Shift Supervision A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by PSL Security Shift Supervision A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by PSL Security Shift Supervision OR                                                                                                                                                                                                                                               CA6 Hazardous event affecting SAFETY SYSTEMS needed for the Security                         Containment                                                 4, 5                                                                                                                                                                                                                                                                                                                                                                                                                               current operating mode OR                                                                       Notification of a credible security threat directed at the site A validated notification from NRC of an aircraft attack                     OR                                                                                                                                                                                                                                               CA6.1                                              5    6 threat within 30 min. of the site                                          A validated notification from the NRC providing information of                                                                                                                                Table C-6     Hazardous Events NOTES                                                                                                                                                                       an aircraft threat                                                                                                                                                                                            None                                The occurrence of any Table C-6 hazardous event Note 1:     The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.                                                                                       [Refer to CA6.1 or SA9.1 for potential escalation                HU2 Seismic event greater than OBE levels                                                6                                                                                              Seismic event (earthquake)
Internal or external FLOODING event AND Event damage has caused indications of degraded 2
performance on one train of a SAFETY SYSTEM needed due to a seismic event]                                                                                                                    Hazardous Note 2:     If an ongoing release is detected and the release start time is unknown, None                                               High winds or tornado strike                              for the current operating mode                                                                             None HU2.1          1        2      3           4         5       6    DEF                   Event                                                                                                                                                       AND EITHER:
assume None that the release duration has exceeded the specified time limit.
None                                                                                              None                                                                                                                         Affecting                                                                                        FIRE Seismic                                                                                                                                                                                                                                                                  Seismic event > OBE as indicated by S-36 CNTMT SEISMIC                                                                                                                                                                                                 Event damage has caused indications of degraded Event                                                                                                                                                                                                                                                                                                                                                           Safety                                                                                          EXPLOSION Note 3:     If the effluent flow past an effluent monitor is known to have stopped due to                                                                                                                               MONITOR HIGH LIMIT alarm                                                                                                                                                                                                                               performance to the second train of the SAFETY Systems                                                                                          Other events with similar hazard characteristics               SYSTEM needed for the current operating mode actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.                                                                                                                                                                                                                                                                                                                                                           as determined by the Shift Manager                             Event damage has resulted in VISIBLE DAMAGE to HU3 Hazardous event                                                                                                                                                                                                                                    the second train of the SAFETY SYSTEM needed for
[Refer to CA6.1 or SA9.1 for potential escalation Note 4:     The pre-calculated effluent monitor values presented in EALs RA1.1,                                                                                   due to a hazardous event]                                                                                                                                                                                                                                                                                    the current operating mode RS1.1 and RG1.1 should be used for emergency classification                                                                                                                                                 HU3.1          1         2     3           4         5       6    DEF                                                                                                                                                                             (Notes 9, 10) assessments until the results from a dose assessment using actual meteorology are available.                                                                                                                                                                                  A tornado strike within the PROTECTED AREA Note 5:     If the equipment in the listed room or area was already inoperable or out-                                                                                                                                  HU3.2          1        2      3           4         5       6    DEF of-service before the event occurred, then no emergency classification is Internal room or area FLOODING of a magnitude sufficient to warranted.
3 require manual or automatic electrical isolation of a SAFETY Note 6:     If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-                                                                                                                                         SYSTEM component required by technical specifications for None time limit, declaration of a General Emergency is not required minute                                                               None                                                                                             None                                  the current operating mode Natural or Tech.                                             Note 7:     This EAL does not apply to routine traffic impediments such as fog, snow,                                                                                                                                   HU3.3          1        2      3           4         5       6    DEF Hazard                                                         ice, or vehicle breakdowns or accidents.                                                                                                                                                                    Movement of personnel within the PROTECTED AREA is Note 8:     A manual trip action is any operator action, or set of actions, which causes                                                                                                                                IMPEDED due to an event external to the PROTECTED the control rods to be rapidly inserted into the core, and does not include                                                                                                                                 AREA involving hazardous materials (e.g., an offsite chemical manually driving in control rods or implementation of boron injection                                                                                                                                       spill or toxic gas release) strategies.                                                                                                                                                                                                 HU3.4          1        2      3           4         5       6    DEF Note 9:     If the affected SAFETY SYSTEM train was already inoperable or out of A hazardous event that results in on-site conditions sufficient service before the hazardous event occurred, then emergency to prohibit the plant staff from accessing the site via personal classification is not warranted.
vehicles (Note 7)
Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY                                                                                                                                          HU4 FIRE potentially degrading the level of safety of the plant
[Refer to CA6.1 or SA9.1 for potential escalation due to fire]
SYSTEM, then this emergency classification is not warranted.
HU4.1          1         2     3           4         5       6    DEF Note 11: One full train of containment heat removal systems consist of any one of the following:                                                                                                                                                                                                 A FIRE is not extinguished within 15 min. of any of the following fire detection indications (Note 1):
Two (2) Containment Spray Headers with flow in each header at least                                                                            Table H-1 PSL Fire Areas                                    Report from the field (i.e., visual observation) 2,700 gpm Receipt of multiple (more than 1) fire zone alarms or Four (4) Containment Coolers                                                                                                               Reactor Containment Building/Shield Building                    indications Two (2) Containment Coolers with at least one (1) Containment Spray                                                                        Control Room                                                    Field verification of a single fire zone alarm Header with at least 2,700 gpm                                                                                                             Reactor Auxiliary Building (RAB)                            AND Diesel Generator Building The FIRE is located in any Table H-1 area Note 12: This threshold is only applicable if all RVLMS Sensors 4 through 8 are not                                                                          Component Cooling Water (CCW) covered inclusive.                                                                                                                                  Intake Structure                                           HU4.2          1         2      3          4         5       6    DEF Intake Cooling Water (ICW) 4 Note 13: If an RCS heat removal system is in operation within the applicable Table                                                                           Steam Trestle Area                                         Receipt of a single fire zone alarm (i.e., no other indications of H                       Fire None not applicable.
None C-4 heat-up duration and RCS temperature is being reduced, the EAL is                                                                               Yard Areas o RWT o CST a FIRE)
AND Hazards                                                                  Note 14: A containment fire zone alarm is considered VALID upon receipt of multiple                                                                          o DOST                                                     The fire zone alarm is indicating a FIRE within any Table H-1 zones (more than one) actuated on CP-376 panel.                                                                                                                                                                area (excluding containment) (Note 14)
AND Note 15: The Intake Structure includes the Intake Cooling Waiter (ICW) system and                                                                                                                                       The existence of a FIRE is not verified (proved or disproved) those immediately surrounding areas that could potentially affect                                                                                                                                              within 30 min. of alarm receipt (Note 1) operability or access to ICW piping and components needed for safe shutdown.                                                                                                                                                                                                      HU4.3          1         2     3           4         5       6    DEF A FIRE within the PROTECTED AREA not extinguished within Table H-2       Safe Operation & Shutdown Rooms/Areas                                                                                                                 60 min. of the initial report, alarm or indication (Note 1)
HU4.4          1         2     3           4         5       6    DEF Room/Area                                Mode A FIRE within the PROTECTED AREA that requires Reactor Auxiliary Building (0.5 and 19.5 levels)            1, 2, 3, 4, 5                                                                                            firefighting support by an offsite fire response agency to Turbine Building (east end levels required for access to    1, 2, 3, 4, 5                                                                                            extinguish AFW, Steam Trestle, Switchgear and Control Rooms)
Cable Spreading Room                                        1, 2, 3, 4, 5                 HA5 Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown A3/B3 Switchgear Rooms                                      1, 2, 3, 4, 5 Steam Trestle Area                                          1, 2, 3, 4, 5                 HA5.1          1        2        3          4     5 5                                                           None Intake Structure (Note 15)
Component Cooling Water Area                 None 1, 2, 3, 4, 5 1, 2, 3, 4, 5 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area                                                                                None Hazardous                                                                                              Emergency Diesel Generator Buildings                        1, 2, 3, 4, 5                   AND Gases                                                                                                                                                                                          Entry into the room or area is prohibited or IMPEDED Refueling Water Tank                                        1, 2, 3, 4, 5 (Note 5)
Water Treatment Plant/Storage Tanks                          1, 2, 3, 4 Containment                                                      4, 5 HS6 Inability to control a key safety function from outside the Control         HA6 Control Room evacuation resulting in transfer of plant control to Room                                                                            alternate locations HS6.1           1        2        3           4    5            6               HA6.1          1       2        3           4    5        6      DEF 6
An event has resulted in the Control Room being evacuated                       An event has resulted in the Control Room being evacuated None AND                                                                                                                                                                                            None Control                                                                                                         Control of any of the following key safety functions is not Room                                                                                                           re-established within 15 min. of the last licensed operator Evacuation                                                                                                       leaving the Control Room (Note 1):
Reactivity (modes 1, 2 and 3 only)
Core cooling RCS heat removal HG7 Other conditions exist which in the judgment of the EC warrant                     HS7 Other conditions existing that in the judgment of the EC warrant            HA7 Other conditions exist that in the judgment of the EC warrant          HU7 Other conditions existing that in the judgment of the EC warrant declaration of a General Emergency                                                     declaration of a Site Area Emergency                                            declaration of an Alert                                                    declaration of an Unusual Event HG7.1               1       2         3         4     5       6           DEF        HS7.1           1        2        3           4    5            6      DEF      HA7.1          1       2        3           4    5        6      DEF       HU7.1          1         2      3          4         5       6    DEF Other conditions exist which, in the judgment of the EC,                            Other conditions exist which, in the judgment of the EC,                         Other conditions exist which, in the judgment of the EC,                   Other conditions exist which, in the judgment of the EC, 7                    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 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 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 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 EC                                                                                                          results in intentional damage or malicious acts, (1) toward site HOSTILE ACTION that results in an actual loss of physical                                                                                                            probable life threatening risk to site personnel or damage to              been initiated. No releases of radioactive material requiring Judgment control of the facility. Releases can be reasonably expected                        personnel or equipment that could lead to the likely failure of                  site equipment because of HOSTILE ACTION. Any                              offsite response or monitoring are expected unless further to exceed EPA Protective Action Guideline exposure levels                           or, (2) that prevent effective access to equipment needed for                    releases are expected to be limited to small fractions of the              degradation of SAFETY SYSTEMS occurs.
offsite for more than the immediate site area.                                      the protection of the public. Any releases are not expected                      EPA Protective Action Guideline exposure levels.
to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.
EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EU1.1        1      2      3       4     5      6     DEF E                        1                                                             None                                                                             None                                                                            None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HSM) > any of the following:
Confinement ISFSI               Boundary                                                                                                                                                                                                                                                                        1,600 mrem/hr on the HSM front bird screen 10 mrem/hr on the outside HSM door centerline 10 mrem/hr on the HSM end shield wall exterior St. Lucie Plant Modes:                                                         1 Power Operation 2
Startup            Hot Standby 3                        4 Hot Shutdown 5
Cold Shutdown 6
Refueling DEF Defueled UNIT 1 Emergency Action Level Matrix                                                              EAL WALL CHART - COLD CONDITIONS (RCS  200&deg;F)
Revision X Prepared for FPL by Operations Support Services, Inc. 4-19-19

PSL2                                               GENERAL EMERGENCY                                                                SITE AREA EMERGENCY                                                                                    ALERT                                                       UNUSUAL EVENT                                                UNIT 2                          GENERAL EMERGENCY                                                        SITE AREA EMERGENCY                                                                              ALERT                                                        UNUSUAL EVENT RG1 Release of gaseous radioactivity resulting in offsite dose greater                RS1 Release of gaseous radioactivity resulting in offsite dose greater         RA1 Release of gaseous or liquid radioactivity resulting in offsite dose   RU1 Release of gaseous or liquid radioactivity greater than 2 times the                                                                                                                                                                                SA1 Loss of all but one AC power source to emergency buses for           SU1 Loss of all offsite AC power capability to emergency buses for 15 SG1 Prolonged loss of all offsite and all onsite AC power to              SS1 Loss of all offsite power and all onsite AC power to emergency than 1,000 mrem TEDE or 5,000 mrem thyroid CDE                                        than 100 mrem TEDE or 500 mrem thyroid CDE                                     greater than 10 mrem TEDE or 50 mrem thyroid CDE                           ODCM limits for 60 minutes or longer                                                                                                                                                                                                                   15 minutes or longer                                                     minutes or longer emergency buses                                                           buses for 15 minutes or longer RG1.1               1       2         3        4      5      6           DEF         RS1.1         1       2        3          4     5           6       DEF       RA1.1         1       2        3           4    5        6      DEF        RU1.1         1       2       3           4         5       6    DEF SG1.1          1          2      3          4                             SS1.1         1       2       3      4                                  SA1.1         1       2       3          4                               SU1.1              1       2        3          4 Reading on any Table 2R-1 effluent radiation monitor
                                                    > column "GE" for  15 min. (Notes 1, 2, 3, 4)
Reading on any Table 2R-1 effluent radiation monitor
                                                                                                                                         > column "SAE" for  15 min. (Notes 1, 2, 3, 4)
Reading on any Table 2R-1 effluent radiation monitor
                                                                                                                                                                                                                       > column "ALERT" for  15 min. (Notes 1, 2, 3, 4)
Reading on any Table 2R-1 effluent radiation monitor
                                                                                                                                                                                                                                                                                                  > column "UE" for  60 min. (Notes 1, 2, 3)                                             1           Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 2A3 and 2B3 Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 2A3 and 2B3 for  15 min. (Note 1)
AC power capability, Table 2S-1, to 4.16 kV emergency buses 2A3 and 2B3 reduced to a single power source for Loss of all offsite AC power capability, Table 2S-1, to 4.16 kV emergency buses 2A3 and 2B3 for  15 min. (Note 1)
Loss of        AND                                                                                                                                               15 min. (Note 1)
RG1.2               1       2         3        4      5      6           DEF         RS1.2         1       2        3          4     5           6       DEF       RA1.2         1       2        3           4    5        6      DEF        RU1.2         1       2       3           4         5       6    DEF                 Emergency      REP CETs > 1200&#xba;F                                                                                                                                   AND Reading on any effluent radiation monitors > 2 x the alarm                           AC Power                                                                                                                                                         Any additional single power source failure will result in loss Dose assessment using actual meteorology indicates                                   Dose assessment using actual meteorology indicates                            Dose assessment using actual meteorology indicates                                                                                                                                                                                                                                                                                                                                                                        Table 2S-1 Unit 2 AC Power Sources doses > 1,000 mrem TEDE or 5,000 mrem thyroid                                        doses > 100 mrem TEDE or 500 mrem thyroid CDE at or                           doses > 10 mrem TEDE or 50 mrem thyroid CDE at or                          setpoint established by a current radioactivity discharge                                                                                                                                                                                             of all AC power to SAFETY SYSTEMS CDE at or beyond the SITE BOUNDARY (Note 4)                                          beyond the SITE BOUNDARY (Note 4)                                             beyond the SITE BOUNDARY (Note 4)                                          permit for  60 min.
(Notes 1, 2, 3)                                                                                                                                                                                                                                                                                                                               Offsite:
1                     RG1.3               1       2         3        4      5      6           DEF         RS1.3         1       2        3          4     5           6       DEF       RA1.3         1       2        3           4    5        6      DEF RU1.3         1       2       3           4         5       6    DEF                                SG2 Loss of all emergency AC and vital DC power sources for 15            SS2 Loss of all vital DC power for 15 minutes or longer 2A Startup Transformer 2B Startup Transformer Rad                     Field survey results indicate EITHER of the following at or                          Field survey results indicate EITHER of the following at or                   Analysis of a liquid effluent sample indicates a                                                                                                                                   minutes or longer Effluent                                                                                                                                                                                                                                                                Sample analysis for a gaseous or liquid release indicates                                                                                                                                                                                                                                                                                     Onsite:
beyond the SITE BOUNDARY:                                                            beyond the SITE BOUNDARY:                                                     concentration or release rate that would result in doses a concentration or release rate 2 x ODCM limits for                                                  SG2.1        1       2       3      4                                    SS2.1         1      2      3     4                                                                                                                               2A Auxiliary Transformer Closed window dose rates > 1,000 mR/hr expected to                                    Closed window dose rates > 100 mR/hr expected to                         > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond continue for  60 min.
Analyses of field survey samples indicate thyroid CDE continue for  60 min.
Analyses of field survey samples indicate thyroid CDE the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) 60 min. (Notes 1, 2) 2           Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 2A3 and 2B3 for  15 min. (Note 1)
Indicated voltage is < 112 VDC on both vital 125 VDC buses 2A and 2B for  15 min. (Note 1) 2B Auxiliary Transformer 2A Diesel Generator 2B Diesel Generator
                                                        > 5,000 mrem for 60 min. of inhalation                                                > 500 mrem for 60 min. of inhalation                                                                                                                                                                                                      Loss of         AND                                                                                                                                                                               None                                                                    None RA1.4         1       2        3           4    5        6      DEF                                                                                                                                                                                                                                                                                                                                                        1AB - 2AB SBO Cross Tie (Notes 1, 2)                                                                         (Notes 1, 2)                                                                                                                                                                                                                                    Vital        Indicated voltage is < 112 VDC on both vital 125 VDC buses                                                                                                                                                                                  2A4 2B4) Cross Tie (if already aligned)
Field survey results indicate EITHER of the following at or                                                                                                     DC Power       2A and 2B for  15 min. (Note 1) beyond the SITE BOUNDARY (Notes 1, 2):
Closed window dose rates > 10 mR/hr expected to continue  60 min.
R                                                                                                                                                                                                                     Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation Table S-2        Safety System Parameters SA3 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress SU3 UNPLANNED loss of Control Room indications for 15 minutes or longer Abnormal Rad Levels
     /
RG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RS2 Spent fuel pool level at the top of the fuel racks                         RA2 Significant lowering of water level above, or damage to, irradiated fuel RU2 UNPLANNED loss of water level above irradiated fuel 3                                            None Reactor power None SA3.1         1          2    3          4 An UNPLANNED event results in the inability to monitor one SU3.1         1        2       3           4 An UNPLANNED event results in the inability to monitor Rad                                               RG2.1             1        2       3          4      5      6           DEF         RS2.1         1       2        3          4      5          6       DEF      RA2.1          1       2        3           4    5        6      DEF       RU2.1          1       2       3           4         5       6    DEF                 Loss of CR                                                                                                   RCS level                                           or more Table S-2 parameters from within the Control Room                    one or more Table S-2 parameters from within the Control Effluent                                                                                                                                                                                                                                                                                                                                                                              Indications                                                                                                                                                      for  15 min. (Note 1)                                                       Room for  15 min. (Note 1)
Spent fuel pool level cannot be restored to at least 37 ft.                         Lowering of spent fuel pool level to 37 ft. 5 in. (Level 3) on               IMMINENT uncovery of irradiated fuel in the REFUELING                      UNPLANNED water level drop in the REFUELING PATHWAY                                                                                                                                                RCS pressure                                          AND 5 in. (Level 3) on LI-04-2A or LI-04-2B for  60 min. (Note 1)                      LI-04-2A or LI-04-2B                                                         PATHWAY                                                                    as indicated by any of the following:
Any significant transient is in progress, Table S-3 SFP or reactor cavity level instrument                                                                                                                                                        CET temperature RA2.2         1    2       3       4       5      6              DEF            SFP or reactor cavity low level alarm SU4 Reactor coolant activity greater than Technical Specification Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds                                                                          Damage to irradiated fuel resulting in a release of                             Report of dropping level in SFP or reactor cavity                                                                                                                                             Level in at least one SG allowable limits 2
radioactivity                                                                AND Release Point & Monitor                       GE                      SAE                 Alert                   UE                                         AND                                                                      UNPLANNED rise in corresponding area radiation levels as                                                                                                                                           Auxiliary feedwater flow to at least                                                                                             SU4.1         1        2        3          4 VALID HIGH alarm on any of the following:                                  indicated by any Table 2R-2 radiation monitors                                                                                                                                                     one SG Plant Vent                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                           Sample analysis indicates that a reactor coolant activity Irradiated Fuel RIM-26-90 (621/622/623)                                                                             2.43E-03 &#xb5;Ci/cc Table 2R-2 area radiation monitors RSC-26-12 Fuel Handling Building Exhaust Vent                                                                                                                   4                                                                                                                                                                                                                                          value is > any of the following allowable limits:
Gaseous
                                                                    - Normal or post SIAS          1.55E+02 &#xb5;Ci/cc      1.55E+01 &#xb5;Ci/cc        1.55E+00 &#xb5;Ci/cc                                                                                                                                                                                                                                                                          None                                                                        None                                                                  None 60 Ci/gm dose equivalent I-131 instantaneous Events                                    - w/ FHB diverted to PV        1.78E+03 &#xb5;Ci/cc      1.78E+02 &#xb5;Ci/cc        1.78E+01 &#xb5;Ci/cc Table 2R-2 Unit 2 Refueling Area Radiation Monitors                                  RCS                                                                                                                                                                                                                                             1.0 Ci/gm dose equivalent I-131 for > 48 hours ECCS Vent                                                                                                                                             RA2.3          1       2        3           4    5        6      DEF                                                                                                                                                                                                                                                                                                                                               518.9 Ci/gm dose equivalent Xe-133 for > 48 hours Activity                                                                                               Table S-3       Significant Transients RIM-26-69 (601/602/603) (A)       7.84E+01 &#xb5;Ci/cc      7.84E+00 &#xb5;Ci/cc        7.84E-01 &#xb5;Ci/cc               N/A RIM-26-70 (611/612/613) (B)                                                                                                                          Lowering of spent fuel pool level to 46 ft. 5 in. (Level 2) on                  ARM #         RMCS Ch.                    Description LI-04-2A or LI-04-2B                                                           RIM-26-2       GAG-002     CCW                                                                                                                                                                 Electrical load rejection > 25% full SG Blowdown Liquid RM-26-5-2A                              N/A                     N/A                 N/A             5.81E-04 &#xb5;Ci/cc                                                                                                                 RIM-26-4       GAG-004     Containment CIS Monitor B                                                                                                                                           electrical load RM-26-6-2B RIM-26-6       GAG-006     Containment CIS Monitor D                                                                                                                                           Reactor Trip                                                                                                                 SU5 RCS leakage for 15 minutes or longer RIM-26-7       GAG-007     Spent Fuel Pool RIM-26-8       GAG-008     Spent Fuel Pool SIAS Activation                                                                                                              SU5.1        1     2      3      4 5
RIM-26-9       GAG-009     Spent Fuel Pool                                                                                                                                                                                                                                                                                  RCS unidentified or pressure boundary leakage RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                 > 10 gpm for  15 min.
RIM-26-10      GAG-010     Spent Fuel Pool Table R-3 Safe Operation & Shutdown Rooms Areas                                                                                                                                                                                                                                                                                                                                       None                                                                     None                                                                  None                                    OR RIM-26-11      GAG-011     Spent Fuel Pool                                           RCS                                                                                                                                                                                                                                    RCS identified leakage > 25 gpm for  15 min.
RA3.1          1       2        3           4    5        6      DEF 3                                               Room/Area                              Mode                                                                                                                                                                            RIM-26-12      GAG-012     Spent Fuel Pool                                         Leakage                                                                                                                                                                                                                                   OR Dose rates > 15 mR/hr in EITHER of the following:                              RIM-26-33      GAG-033     Refueling Canal                                                                                                                                                                                                                                                                                  Leakage from the RCS to a location outside containment Reactor Auxiliary Building (0.5 and 19.5 levels)       1, 2, 3, 4, 5 S
None                                                                              None                                           Control Room                                                                                              None                                                                                                                                                                                                                                                                                        > 25 gpm for  15 min.
Area                           Turbine Building (east end levels required for                                                                                                                    Central Alarm Station                                                                                                                                                                                                                                                                                                                                                                                 (Note 1)
Radiation                        access to AFW, Steam Trestle, Switchgear and           1, 2, 3, 4, 5 Control Rooms)                                                                                                                                               RA3.2          1       2        3           4    5 Levels                                                                                                                                                                                                                                                                                                                                            System                                                                                               SS6 Inability to shut down the reactor causing a challenge to core      SA6 Automatic or manual trip fails to shut down the reactor and          SU6 Automatic or manual trip fails to shut down the reactor Cable Spreading Room                                   1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                       cooling or RCS heat removal                                             subsequent manual actions taken at the reactor control consoles An UNPLANNED event results in radiation levels that                                                                                                  Malfunct.
are not successful in shutting down the reactor A3/B3 Switchgear Rooms                                 1, 2, 3, 4, 5                                                                                         prohibit or IMPEDE access to any Table R-3 room or Steam Trestle Area                                     1, 2, 3, 4, 5                                                                                         area (Note 5)                                                                                                                                                                                                                                             SS6.1          1        2                                                SA6.1        1          2                                                   SU6.1         1        2 Intake Structure (Note 15)                             1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                   An automatic or manual trip did not shut down the reactor as            An automatic or manual trip (RX Trip Pushbuttons) did not                An automatic trip did not shut down the reactor as HS1 HOSTILE ACTION within the PROTECTED AREA                                   HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or                     HU1 Confirmed SECURITY CONDITION or threat Component Cooling Water Area                           1, 2, 3, 4, 5                                                                                             airborne attack threat within 30 minutes                                                                                                                                                                                                              indicated by reactivity control Safety Function Status Check            shut down the reactor as indicated by reactivity control                 indicated by reactivity control Safety Function Status Emergency Diesel Generator Buildings                   1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                   acceptance criteria not met                                             Safety Function Status Check acceptance criteria not met                 Check acceptance criteria not met after any RPS HS1.1           1       2        3          4     5           6       DEF      HA1.1          1       2        3           4    5        6      DEF       HU1.1          1       2        3          4         5       6    DEF Refueling Water Tank                                   1, 2, 3, 4, 5 AND                                                                     AND                                                                    setpoint is exceeded 1                                                                                                                                                                                                                                                                                                                                                           6 A HOSTILE ACTION is occurring or has occurred within the                       A HOSTILE ACTION is occurring or has occurred within                       A SECURITY CONDITION that does not involve a HOSTILE                                                                                      None                                 All actions taken to shut down the reactor are not successful           Subsequent automatic or manual trip actions (RX Trip                       AND Water Treatment Plant/Storage Tanks                     1, 2, 3, 4                                                                                                                                                                                                                                                                                                                                                     as indicated by reactivity control Safety Function Status               Pushbuttons) are not successful in shutting down the                     A subsequent automatic trip or manual trip (RX Trip None                                         PROTECTED AREA as reported by PSL Security Shift                               the OWNER CONTROLLED AREA as reported by PSL                               ACTION as reported by PSL Security Shift Supervision Containment                                                  4, 5             Supervision                                                                    Security Shift Supervision                                                  OR                                                                                                                                                                            Check acceptance criteria not met                                       reactor as indicated by reactivity control Safety Function               Pushbuttons) are successful in shutting down the reactor Security                                                                                                                                                                                       OR                                                                       Notification of a credible security threat directed at the site                          RPS                                                                                     AND EITHER:                                                           Status Check acceptance criteria not met (Note 8)                        as indicated by reactivity control Safety Function Status A validated notification from NRC of an aircraft attack                     OR                                                                                     Failure                                                                                                                                                                                                                                 Check acceptance criteria met (Note 8)
REP CETs >1200&deg;F threat within 30 min. of the site                                          A validated notification from the NRC providing information of                                                                                                                                                                                                                                                                   SU6.2        1      2 NOTES                                                                                                                                                                     an aircraft threat                                                                                                                                                                  Applicable RCS and Core Heat Removal (HR) Safety Function Status Check acceptance criteria not met                                                                                               A manual trip (RX Trip Pushbuttons) did not shut down the Note 1:   The EC should declare the event promptly upon determining that the time                                                                                                                                                                                                                                                                                                                                                                                                                                                                        reactor as indicated by reactivity control Safety Function limit has been exceeded, or will likely be exceeded.                                                                                      [Refer to CA6.1 or SA9.1 for potential escalation                HU2 Seismic event greater than OBE levels                                                                                                                                                                                                                                                                                           Status Check acceptance criteria not met (Note 8) 2                                             Note 2:   If an ongoing release is detected and the release start time is unknown, assume None that the release duration has exceeded the specified time limit.
None due to a seismic event]
None HU2.1          1       2       3           4         5       6    DEF AND A subsequent automatic trip or manual trip action (RX Trip Pushbuttons) is successful in shutting down the reactor as Seismic                                                                                                                                                                                                                                                                Seismic event > OBE as indicated by S-44 CNTMT SEISMIC Note 3:   If the effluent flow past an effluent monitor is known to have stopped due to                                                                                                                                                                                                                                                                                                                                                                                                                                                                  indicated by reactivity control Safety Function Status Check Event                                                                                                                                                                                                                                                                 MONITOR HIGH LIMIT alarm actions to isolate the release path, then the effluent monitor reading is no                                                                                                                                                                                                                                                                                                                                                                                                                                                                   acceptance criteria met Table S-4      Communications Methods longer VALID for classification purposes.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      SU7 Loss of all onsite or offsite communications capabilities HU3 Hazardous event
[Refer to CA6.1 or SA9.1 for potential escalation Note 4:   The pre-calculated effluent monitor values presented in EALs RA1.1,                                                                                                                                                                                                                                                                                                                                                                                                      System                     Onsite     Offsite     NRC 7
due to a hazardous event]                                                                                                                                                                                                                                                                                                                                                                 SU7.1        1      2      3      4 RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual                                                                                                                                          HU3.1          1       2       3           4         5       6    DEF                                                                                                                                                                    Plant Radio System                                 X                                       Loss of all Table S-4 onsite communication methods None                                                                     None                                                                  None meteorology are available.                                                                                                                                                                                 A tornado strike within the PROTECTED AREA                                                                                                                                                                                               Commercial Telephone System                        X          X          X                  OR Loss of Loss of all Table S-4 State and local agency communication Commun.                                                                                                                                            Plant Address (PA) System                          X Note 5:   If the equipment in the listed room or area was already inoperable or out-                                                                                                                                 HU3.2          1       2       3           4         5       6    DEF                                                                                                                                                                                                                                                               methods of-service before the event occurred, then no emergency classification is                                                                                                                                                                                                                                                                                                                                                                           Federal Telephone System (ENS)                                           X                  OR Internal room or area FLOODING of a magnitude sufficient to warranted.                                                                                                                                                                                                                                                                                                                                                                                                                                          EMnet                                                         X                            Loss of all Table S-4 NRC communication methods require manual or automatic electrical isolation of a SAFETY 3                                             Note 6:   If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute None time limit, declaration of a General Emergency is not required None                                                                                            None SYSTEM component required by technical specifications for the current operating mode Hot Ring Down (HRD) Telephone System Shift Manager Cell Phones X
X          X SU8 Failure to isolate containment or loss of containment pressure control Natural or                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                SU8.1             1       2        3           4 8
Tech.                                              Note 7:   This EAL does not apply to routine traffic impediments such as fog, snow,                                                                                                                                  HU3.3          1       2       3           4         5       6    DEF Hazard                                                        ice, or vehicle breakdowns or accidents.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                   Any penetration is not closed within 15 min. of a VALID Movement of personnel within the PROTECTED AREA is None                                                                     None                                                                  None                                   CIAS actuation signal Note 8:   A manual trip action is any operator action, or set of actions, which causes                                                                                                                               IMPEDED due to an event external to the PROTECTED CTMT                                                                                                                                                                                                                                      OR the control rods to be rapidly inserted into the core, and does not include                                                                                                                                AREA involving hazardous materials (e.g., an offsite chemical Failure                                                                                                                                                                                                                                 CTMT pressure > 5.4 psig with < one full train of CTMT manually driving in control rods or implementation of boron injection                                                                                                                                      spill or toxic gas release) heat removal systems operating per design for  15 min.
strategies.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                (Notes 1, 11)
HU3.4          1       2       3           4         5       6    DEF Note 9:   If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency                                                                                                                                                A hazardous event that results in on-site conditions sufficient classification is not warranted.                                                                                                                                                                           to prohibit the plant staff from accessing the site via personal                                                                                                                                                                                        SA9 Hazardous event affecting SAFETY SYSTEMS needed for the vehicles (Note 7)                                                                                                                                                                                                                                           current operating mode Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no                                                                                                                                                                                                                                                                                                                                               Table S-5    Hazardous Events indications of degraded performance to at least one train of a SAFETY                                                                                                                                       HU4 FIRE potentially degrading the level of safety of the plant                                                                                                                                                                                         SA9.1        1     2        3    4 SYSTEM, then this emergency classification is not warranted.                                                                         [Refer to CA6.1 or SA9.1 for potential escalation due to fire]
Seismic event (earthquake)                                    The occurrence of any Table S-5 hazardous event Note 11: One full train of containment heat removal systems consist of any one of the following:
HU4.1          1       2       3           4         5 A FIRE is not extinguished within 15 min. of any of the 6    DEF 9                                                                                               Internal or external FLOODING event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed Two (2) Containment Spray Headers with flow in each header at least                                                                                                                                following fire detection indications (Note 1):                                        Hazardous Table H-1 PSL Fire Areas                                    Report from the field (i.e., visual observation)
None High winds or tornado strike                                  for the current operating mode                                                                                  None 2,700 gpm                                                                                                                                                                                                                                                                                  Event Receipt of multiple (more than 1) fire zone alarms or                                                                                                                                                                                                AND EITHER:
Four (4) Containment Coolers                                                                                                                                                                                                                                                              Affecting                                                                                          FIRE Reactor Containment Building/Shield Building                    indications                                                                                                                                                                                                                                             Event damage has caused indications of degraded Safety Two (2) Containment Coolers with at least one (1) Containment Spray                                                                     Control Room                                                    Field verification of a single fire zone alarm                                                                                                                                                                                                          performance to the second train of the SAFETY Systems                                                                                            EXPLOSION Header with at least 2,700 gpm                                                                                                          Reactor Auxiliary Building (RAB)                            AND                                                                                                                                                                                                                                                         SYSTEM needed for the current operating mode Diesel Generator Building                                                                                                                                                                                                                                                                                               Event damage has resulted in VISIBLE DAMAGE to Note 12: This threshold is only applicable if all RVLMS Sensors 4 through 8 are not                                                                       Component Cooling Water (CCW)
The FIRE is located in any Table H-1 area                                                                                                                                                 Other events with similar hazard characteristics covered inclusive.                                                                                                                                                                                                                                                                                                                                                                                                                                                       the second train of the SAFETY SYSTEM needed for Intake Structure                                                                                                                                                                                                                                     as determined by the Shift Manager Intake Cooling Water (ICW)
HU4.2          1       2        3          4         5       6    DEF                                                                                                                                                                                        the current operating mode Note 13: If an RCS heat removal system is in operation within the applicable Table                                                                                                                                                                                                                                                                                                                                                                                           (Notes 9, 10) 4 Steam Trestle Area                                         Receipt of a single fire zone alarm (i.e., no other indications of H                                                                              C-4None heat-up duration and RCS temperature is being reduced, the EAL not applicable.
None is                                                                       Yard Areas a FIRE)
F o RWT o CST                                                        AND                                                                                                 FG1.1         1          2      3          4                              FS1.1         1        2       3          4                             FA1.1        1          2    3          4 Fire                                          Note 14: A containment fire zone alarm is considered VALID upon receipt of multiple                                                                                                                                  The fire zone alarm is indicating a FIRE within any Table H-1 Hazards                                                                                                                                                                                                                             o DOST                                                                                                                                                           Loss of any two barriers                                                  Loss or potential loss of any two barriers (Table F-1)                  Any loss or any potential loss of EITHER Fuel Clad or zones (more than one) actuated on CP-376 panel.                                                                                                                                                             area (excluding containment) (Note 14)
None Fission Product              AND                                                                                                                                             RCS (Table F-1)
Note 15: The Intake Structure includes the Intake Cooling Waiter (ICW) system and                                                                                                                                      AND                                                                                                 Loss or potential loss of the third barrier (Table F-1)
Barrier Degradation those immediately surrounding areas that could potentially affect                                                                                                                                           The existence of a FIRE is not verified (proved or disproved) operability or access to ICW piping and components needed for safe                                                                                                                                          within 30 min. of alarm receipt (Note 1) shutdown.                                                                                                                                                                                                   HU4.3          1       2       3           4         5       6    DEF A FIRE within the PROTECTED AREA not extinguished within Table H-2      Safe Operation & Shutdown Rooms/Areas                                                                                                                60 min. of the initial report, alarm or indication (Note 1)
Room/Area                                Mode HU4.4          1       2       3           4         5       6    DEF Table F-1 Fission Product Barrier Matrix A FIRE within the PROTECTED AREA that requires Reactor Auxiliary Building (0.5 and 19.5 levels)           1, 2, 3, 4, 5                                                                                            firefighting support by an offsite fire response agency to Turbine Building (east end levels required for access to   1, 2, 3, 4, 5                                                                                            extinguish                                                                                                                         Fuel Clad (FC) Barrier                                                           Reactor Coolant System (RCS) Barrier                                                                     Containment (CTMT) Barrier AFW, Steam Trestle, Switchgear and Control Rooms)
Cable Spreading Room                                       1, 2, 3, 4, 5                 HA5 Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown                                                                                                       Category                                Loss                                     Potential Loss                                              Loss                                Potential Loss                                             Loss                                           Potential Loss A3/B3 Switchgear Rooms                                     1, 2, 3, 4, 5 Steam Trestle Area                                         1, 2, 3, 4, 5                 HA5.1          1        2        3          4     5                                                                                                                                                                                                                               1. An automatic or manual SIAS                1. UNISOLABLE RCS or SG tube                      1. A leaking or RUPTURED SG is 5
: 1. RVLMS indicates Sensors 4 through 8                                                             leakage > 50 gpm Intake Structure (Note 15)                                 1, 2, 3, 4, 5                 Release of a toxic, corrosive, asphyxiant or flammable gas                                                                                                                                                                                                                           actuation required by EITHER:                                                                   FAULTED outside of CTMT not covered (Note 12)
Component Cooling Water Area                   None        1, 2, 3, 4, 5                 into any Table H-2 room or area                                                                                                                                                                                                                                                       UNISOLABLE RCS leakage A
None                                                                                                                                                                                                                                       None Hazardous                                                                                             Emergency Diesel Generator Buildings                       1, 2, 3, 4, 5                   AND                                                                                                                                                                                                                                                                                 SG tube RUPTURE                           2. Uncontrolled RCS cooldown None                                                                                                                                                                                                                                                       None Gases                                                                                                                                                                                        Entry into the room or area is prohibited or IMPEDED                                                                                                   RCS or SG Tube                                                                                                                                                              > 100&#xba;F/hr and RCS pressure and Refueling Water Tank                                       1, 2, 3, 4, 5 (Note 5)                                                                                                                                                  Leakage                                                                                                                                                                  temperature are to the left of the Water Treatment Plant/Storage Tanks                         1, 2, 3, 4 Maximum Subcooled (PT) Curve Containment                                                      4, 5                                                                                                                                                                                                                                                                                                                                                       (EOP Figure 1A or 1B)
HS6 Inability to control a key safety function from outside the Control        HA6 Control Room evacuation resulting in transfer of plant control to                                                                                                          1. REP CETs > 1200&deg;F                             2. REP CETs > 700&deg;F                                                                             3. Both SG wide range level < 15% for                                                                     1. REP CETs > 1200&deg;F Room                                                                           alternate locations                                                                                                                                                                                                                                                                                                             15 min.                                                                                                   AND HS6.1           1         2       3         4      5           6               HA6.1          1       2        3           4    5        6      DEF                                                                                         B                                                                  3. Both SG wide range level < 15% for 15 min.                                                             None AND OTC not established                                                  None Functional Recovery (2-EOP-15) for RCS and Core Heat Removal Inadequate Heat 6                                                                                                       An event has resulted in the Control Room being evacuated                      An event has resulted in the Control Room being                                                                                                                                                                                     AND                                                                                                                                                                                                      not effective within 15 min.
evacuated                                                                                                                                                  Removal                                                                 OTC not established                                                                                                                                                                                       (Note 1)
None                                         AND                                                                                                                                                                                           None Control                                                                                                        Control of any of the following key safety functions is not Room                                                                                                          re-established within 15 min. of the last licensed operator Evacuation                                                                                                      leaving the Control Room (Note 1):                                                                                                                                                                                                                            2. CTMT high range radiation                                                                       2. Any CIS monitor RIM-26-3/4/5/6                                                                                                                       2. CTMT high range radiation Reactivity (modes 1, 2 and 3 only)                                                                                                                                                                                                                           monitors RIM-26-40/41                                                                              > 1.5E+03 mR/hr                                                                                                                                         monitors RIM-26-40/41 Core cooling                                                                                                                                                                                                                                                 > 1.4E+02 R/hr                                                                                                                                                                                                                                             > 2.7E+03 R/hr RCS heat removal C                 3. Coolant activity > 300 Ci/gm dose equivalent I-131 CTMT Radiation /
None                                                                                             None                                                None HG7 Other conditions exist which in the judgment of the EC warrant                    HS7 Other conditions existing that in the judgment of the EC warrant           HA7 Other conditions exist that in the judgment of the EC warrant          HU7 Other conditions existing that in the judgment of the EC warrant          RCS Activity declaration of a General Emergency                                                    declaration of a Site Area Emergency                                           declaration of an Alert                                                    declaration of an Unusual Event HG7.1                1       2       3          4      5      6           DEF       HS7.1           1         2       3         4      5           6      DEF      HA7.1          1       2        3           4    5        6      DEF       HU7.1          1       2        3          4         5       6    DEF Other conditions exist which, in the judgment of the EC,                            Other conditions exist which, in the judgment of the EC,                       Other conditions exist which, in the judgment of the EC,                   Other conditions exist which, in the judgment of the EC, 7                    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 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 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 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
: 2. CTMT isolation (CIAS) is required AND EITHER:
: 3. CTMT pressure > 42 psig EC                                                                                                          results in intentional damage or malicious acts, (1) toward site HOSTILE ACTION that results in an actual loss of physical                                                                                                          probable life threatening risk to site personnel or damage to              been initiated. No releases of radioactive material requiring                                                                                                                                                                                                                                                  CTMT integrity has been lost                    4. CTMT hydrogen concentration Judgment control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.                                             D                                                                                                                                                                                                                            based on EC judgment                               > 4%
CTMT Integrity or                           None                                            None                                                  None                                       None                                 UNISOLABLE pathway from                          5. CTMT pressure > 5.4 psig with offsite for more than the immediate site area.                                      the protection of the public. Any releases are not expected                    EPA Protective Action Guideline exposure levels.
Bypass                                                                                                                                                                                                                         CTMT atmosphere to the                              < one full train of CTMT heat to result in exposure levels which exceed EPA Protective environment exists                                  removal systems operating per Action Guideline exposure levels beyond the SITE BOUNDARY.                                                                                                                                                                                                                                                                                                                                                                                                                                                        3. Indications of RCS leakage outside                      design for  15 min.
of CTMT                                                 (Notes 1, 11)
EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EU1.1        1      2      3       4     5      6     DEF                                           4. Any condition in the opinion of the           4. Any condition in the opinion of the EC         3. Any condition in the opinion of the        4. Any condition in the opinion of the            4. Any condition in the opinion of the                  6. Any condition in the opinion of the E
E                       1                                                             None                                                                            None                                                                           None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HSM) > any of the following:
EC Judgment EC that indicates loss of the Fuel Clad barrier that indicates potential loss of the Fuel Clad barrier EC that indicates loss of the RCS barrier EC that indicates potential loss of the RCS barrier EC that indicates loss of the CTMT barrier EC that indicates potential loss of the CTMT barrier Confinement ISFSI              Boundary                                                                                                                                                                                                                                                                      1,600 mrem/hr on the HSM front bird screen 10 mrem/hr on the outside HSM door centerline 10 mrem/hr on the HSM end shield wall exterior St. Lucie Plant Modes:                                                         1 Power Operation 2
Startup 3
Hot Standby 4
Hot Shutdown 5
Cold Shutdown 6
Refueling DEF Defueled UNIT 2 Emergency Action Level Matrix                                                                 EAL WALL CHART - HOT CONDITIONS (RCS > 200&deg;F)
Revision X Prepared for FPL by Operations Support Services, Inc. 4-19-19

PSL2                                                GENERAL EMERGENCY                                                                     SITE AREA EMERGENCY                                                                                  ALERT                                                        UNUSUAL EVENT                                               UNIT 2                        GENERAL EMERGENCY                                                         SITE AREA EMERGENCY                                                                        ALERT                                                       UNUSUAL EVENT RG1 Release of gaseous radioactivity resulting in offsite dose greater                    RS1 Release of gaseous radioactivity resulting in offsite dose greater        RA1 Release of gaseous or liquid radioactivity resulting in offsite dose   RU1 Release of gaseous or liquid radioactivity greater than 2 times the                            CG1 Loss of RCS inventory affecting fuel clad integrity with               CS1 Loss of RCS inventory affecting core decay heat removal         CA1 Significant loss of RCS inventory                                     CU1 UNPLANNED loss of RCS inventory than 1,000 mrem TEDE or 5,000 mrem thyroid CDE                                            than 100 mrem TEDE or 500 mrem thyroid CDE                                    greater than 10 mrem TEDE or 50 mrem thyroid CDE                           ODCM limits for 60 minutes or longer                                                               containment challenged                                                     capability RG1.1              1       2        3          4         5      6            DEF         RS1.1         1        2        3          4     5          6      DEF       RA1.1         1       2        3          4     5        6      DEF        RU1.1         1       2       3           4      5      6        DEF CG1.1                                             5       6                CS1.1                                        5       6              CA1.1                                              5       6              CU1.1                                           5        6 Reading on any Table 2R-1 effluent radiation monitor                                     Reading on any Table 2R-1 effluent radiation monitor                         Reading on any Table 2R-1 effluent radiation monitor                       Reading on any Table 2R-1 effluent radiation monitor
                                                    > column "GE" for  15 min. (Notes 1, 2, 3, 4)                                           > column "SAE" for  15 min. (Notes 1, 2, 3, 4)                              > column "ALERT" for  15 min. (Notes 1, 2, 3, 4)                          > column "UE" for  60 min. (Notes 1, 2, 3)                                                       RCS level cannot be monitored for  30 min. (Note 1)                       0 in. on LI-1117 or LI-1117-1                                       Loss of RCS inventory as indicated by RCS water level                     UNPLANNED loss of reactor coolant results in RCS water AND                                                                                                                                           < minimum required for continued LPSI pump operation                      level less than a required lower limit for  15 min. (Note 1) 1       2       3           4      5      6        DEF                              Core uncovery is indicated by any of the following:                        CS1.2                                        5       6 RG1.2              1       2        3          4         5      6            DEF         RS1.2         1        2        3          4     5          6      DEF       RA1.2         1       2        3          4     5        6      DEF        RU1.2 UNPLANNED increase in any Table C-1 sump or tank                                                                                             CA1.2                                             5       6              CU1.2                                           5        6 RCS level cannot be monitored for  30 min. (Note 1) 1 Dose assessment using actual meteorology indicates                                       Dose assessment using actual meteorology indicates                           Dose assessment using actual meteorology indicates                         Reading on any effluent radiation monitors > 2 x the alarm                                           level of sufficient magnitude to indicate core uncovery doses > 1,000 mrem TEDE or 5,000 mrem thyroid                                            doses > 100 mrem TEDE or 500 mrem thyroid CDE at or                          doses > 10 mrem TEDE or 50 mrem thyroid CDE at or                          setpoint established by a current radioactivity discharge                                                                                                                     AND Visual observation of UNISOLABLE RCS leakage of                                                                                             RCS water level cannot be monitored for  15 min. (Note 1)                RCS level cannot be monitored CDE at or beyond the SITE BOUNDARY (Note 4)                                              beyond the SITE BOUNDARY (Note 4)                                            beyond the SITE BOUNDARY (Note 4)                                          permit for  60 min.                                                                                                                                                         Core uncovery is indicated by any of the following:
sufficient magnitude to indicate core uncovery                                                                                               AND EITHER                                                                AND EITHER:
(Notes 1, 2, 3)                                                                      RCS Level                                                                                  UNPLANNED increase in any Table C-1 sump or tank 1                     RG1.3              1       2        3          4         5      6            DEF         RS1.3         1        2        3          4     5          6      DEF       RA1.3         1       2        3          4     5        6      DEF RU1.3         1       2       3           4      5      6        DEF Containment High Range Radiation Monitor reading
                                                                                                                                                                                                                                                                                                                                                                                                          > 4.8E+04 R/hr (Refueling Mode)
Erratic source range monitor indications level of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage UNPLANNED increase in any Table C-1 sump or tank level due to a loss of RCS inventory Visual observation of UNISOLABLE RCS leakage Rad                     Field survey results indicate EITHER of the following at or                              Field survey results indicate EITHER of the following at or                  Analysis of a liquid effluent sample indicates a                                                                                                                                                                                                           sufficient magnitude to indicate core uncovery Sample analysis for a gaseous or liquid release indicates                                          AND Effluent                  beyond the SITE BOUNDARY:                                                                beyond the SITE BOUNDARY:                                                    concentration or release rate that would result in doses                                                                                                                                                                                                   Containment High Range Radiation Monitor reading a concentration or release rate 2 x ODCM limits for                                               Any Containment Challenge indication, Table C-2 Closed window dose rates > 1,000 mR/hr expected to                                      Closed window dose rates > 100 mR/hr expected to                         > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond                                                                                                                                                                                                         > 4.8E+04 R/hr (Refueling Mode) continue for  60 min.                                                                  continue for  60 min.                                                   the SITE BOUNDARY for 60 min. of exposure                                   60 min. (Notes 1, 2)                                                                                                                                                          Erratic source range monitor indications Analyses of field survey samples indicate thyroid CDE                                   Analyses of field survey samples indicate thyroid CDE                    (Notes 1, 2)
                                                         > 5,000 mrem for 60 min. of inhalation                                                  > 500 mrem for 60 min. of inhalation                                                                                                                                                                                                                        Table C-2      Containment Challenge Indications                                        Table C-1      Sumps/Tanks (Notes 1, 2)                                                                             (Notes 1, 2)                                                                 RA1.4         1       2        3          4     5        6      DEF Field survey results indicate EITHER of the following at or                                                                                                                             CONTAINMENT CLOSURE not established                                                     Containment Sump                            CA2 Loss of all offsite and all onsite AC power to emergency buses for    CU2 Loss of all but one AC power source to emergency buses for 15 beyond the SITE BOUNDARY (Notes 1, 2):                                                                                                                                                  (Note 6)                                                                                                                                15 minutes or longer                                                      minutes or longer Reactor Cavity Sump Closed window dose rates > 10 mR/hr expected to continue  60 min.                                                                                                                                                                  CTMT hydrogen concentration > 4%                                                        Safeguards Sump                             CA2.1                                              5       6        DEF   CU2.1                                           5        6      DEF R                                                                                                                                                                                                                        Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation UNPLANNED increase in CTMT pressure None Holdup Tanks None Refueling Water Tank Loss of all offsite and all onsite AC power to 4.16 kV emergency buses 2A3 and 2B3 for  15 min. (Note 1)
AC power capability, Table 1C-3, to 4.16 kV emergency buses 2A3 and 2B3 reduced to a single power source for Abnormal Rad Levels RG2 Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RS2 Spent fuel pool level at the top of the fuel racks                        RA2 Significant lowering of water level above, or damage to, irradiated fuel RU2 UNPLANNED loss of water level above irradiated fuel                                  2                     Table 2C-3 Unit 2 AC Power Sources CCW Surge Tank Quench Tank 15 min. (Note 1)
AND Loss of                                                                                                                                                                                                                              Any additional single power source failure will result in
     /                                                                                                                                                                                                                                                                                                                                                                                    Emergency Equipment Drain Tank loss of all AC power to SAFETY SYSTEMS Rad                                               RG2.1             1       2        3          4        5       6        DEF             RS2.1         1        2        3          4     5          6      DEF      RA2.1          1       2        3          4     5        6      DEF       RU2.1          1       2       3           4      5      6        DEF                 AC Power                 Offsite:
Effluent                                            Spent fuel pool level cannot be restored to at least 37 ft.                             Lowering of spent fuel pool level to 37 ft. 5 in. (Level 3) on              IMMINENT uncovery of irradiated fuel in the REFUELING                      UNPLANNED water level drop in the REFUELING PATHWAY                                                               2A Startup Transformer 5 in. (Level 3) on LI-04-2A or LI-04-2B for  60 min. (Note 1)                          LI-04-2A or LI-04-2B                                                        PATHWAY                                                                    as indicated by any of the following:                                                                             2B Startup Transformer SFP or reactor cavity level instrument                                                                       2A Auxiliary Transformer back fed via RA2.2         1    2       3       4       5      6              DEF            SFP or reactor cavity low level alarm                                                                        Main Transformer (if already aligned)
Table 2R-1 Unit 2 Effluent Monitor Classification Thresholds                                                                             Damage to irradiated fuel resulting in a release of                             Report of dropping level in SFP ot reactor cavity                                                            2B Auxiliary Transformer back fed via Main Transformer (if already aligned) 2 radioactivity                                                                AND Release Point & Monitor                        GE                       SAE                  Alert                    UE                                      AND                                                                      UNPLANNED rise in corresponding area radiation levels as                                                                                                                                                                                          CA3 Inability to maintain plant in cold shutdown                          CU3 UNPLANNED increase in RCS temperature VALID HIGH alarm on any of the following:                                  indicated by any Table 2R-2 radiation monitors                                                                 Onsite:
Plant Vent 2A Diesel Generator C
Irradiated                                RIM-26-90 (621/622/623)                                                                                 2.43E-03 &#xb5;Ci/cc Table 2R-2 area radiation monitors Fuel                                                                                                                                                                                              RSC-26-12 Fuel Handling Building Exhaust Vent                                                                                                                                           2B Diesel Generator                                                                                                              CA3.1                                              5       6              CU3.1                                           5        6 Gaseous
                                                                    - Normal or post SIAS              1.55E+02 &#xb5;Ci/cc      1.55E+01 &#xb5;Ci/cc         1.55E+00 &#xb5;Ci/cc Events                                    - w/ FHB diverted to PV ECCS Vent 1.78E+03 &#xb5;Ci/cc      1.78E+02 &#xb5;Ci/cc         1.78E+01 &#xb5;Ci/cc RA2.3          1       2        3          4     5        6      DEF Table 2R-2 Unit 2 Refueling Area Radiation Monitors Cold S/D
                                                                                                                                                                                                                                                                                                                                                                              / Refuel 3                       1AB - 2AB SBO Cross Tie 2A4 2B4) Cross  Tie (if already aligned)
None UNPLANNED increase in RCS temperature to > 200&deg;F for
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      > Table C-4 duration due to a loss of RCS cooling UNPLANNED increase in RCS temperature to > 200&deg;F due to loss of decay heat removal capability RIM-26-69 (601/602/603) (A)          7.84E+01 &#xb5;Ci/cc      7.84E+00 &#xb5;Ci/cc         7.84E-01 &#xb5;Ci/cc               N/A                                                                                                                                                                                                        RCS                                                                                                                                                      (Notes 1, 13)
RIM-26-70 (611/612/613) (B)                                                                                                                            Lowering of spent fuel pool level to 46 ft. 5 in. (Level 2) on                 ARM #          RMCS Ch.                Description                    System        Temp.                                                                                                                                                       OR                                                                      CU3.2                                           5       6 LI-04-2A or LI-04-2B                                                          RIM-26-2        GAG-002    CCW                                         Malfunct.                                                                                                                                                               UNPLANNED RCS pressure rise > 10 psig SG Blowdown                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    Loss of all RCS temperature and RCS water level indication Liquid RM-26-5-2A                                N/A                      N/A                   N/A            5.81E-04 &#xb5;Ci/cc                                                                                                              RIM-26-4        GAG-004    Containment CIS Monitor B                                                                                                                                                                                           (does not apply to solid plant conditions)                                for  15 min. (Note 1)
RM-26-6-2B                                                                                                                                                                                                                                                                                                                                                                    Table C-4 RCS Heatup Duration Thresholds RIM-26-6        GAG-006    Containment CIS Monitor D RIM-26-7        GAG-007    Spent Fuel Pool                                                                                                                                                                                                                                                                               CU4 Loss of vital DC power for 15 minutes or longer CONTAINMENT CLOSURE RIM-26-8        GAG-008    Spent Fuel Pool                                                                                 RCS Status                                               Heat-up Duration Status                                                                                                                                                           CU4.1                                           5        6 Table R-3 Safe Operation & Shutdown Rooms Areas RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown RIM-26-9 RIM-26-10 GAG-009 GAG-010 Spent Fuel Pool Spent Fuel Pool                                            4                                   Intact AND not reducedNone inventory N/A                        60 minutes       None Indicated voltage is < 112 VDC on required vital 125 VDC buses 2A and 2B for  15 min. (Note 1)
RIM-26-11       GAG-011    Spent Fuel Pool                                          Loss of RA3.1          1       2        3          4     5        6      DEF                                                                                              Vital DC                                                                                              20 minutes 3
Room/Area                                   Mode                                                                                                                                                                          RIM-26-12       GAG-012    Spent Fuel Pool                                                                                                                 Established Power                                Not intact OR                                                                                                        Table C-5      Communications Methods Dose rates > 15 mR/hr in EITHER of the following:                             RIM-26-33       GAG-033    Refueling Canal Reactor Auxiliary Building (0.5 and 19.5 levels)          1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                              reduced inventory None                                                                                    None                                         Control Room                                                                                              None Not Established                   0 minutes Area                            Turbine Building (east end levels required for Central Alarm Station access to AFW, Steam Trestle, Switchgear and              1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                                                                                            System                      Onsite     Offsite          NRC Radiation Control Rooms)                                                                                                                                                 RA3.2          1       2        3          4     5                                                                                                                                                                                                                                                                                                                                                     CU5 Loss of all onsite or offsite communications capabilities Levels                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         Plant Radio System                                    X Cable Spreading Room                                      1, 2, 3, 4, 5 An UNPLANNED event results in radiation levels that                                                                                                                                                                                                                                                            Commercial Telephone System                           X              X        X         CU5.1 A3/B3 Switchgear Rooms                                    1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                               5       6       DEF 5
prohibit or IMPEDE access to any Table R-3 room or                                                                                                                                                                                                                                                             Plant Address (PA) System                             X Steam Trestle Area                                        1, 2, 3, 4, 5                                                                                        area (Note 5)                                                                                                                                                                                                                                                                                                                                                                                          Loss of all Table C-5 onsite communication methods None                                                                     None                  Federal Telephone System (ENS)                                                X Intake Structure (Note 15)                                1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                OR HS1 HOSTILE ACTION within the PROTECTED AREA                                  HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or                     HU1 Confirmed SECURITY CONDITION or threat                                             Loss of                                                                                                                                      EMnet                                                                X                  Loss of all Table C-5 State and local agency communication Component Cooling Water Area                              1, 2, 3, 4, 5                                                                                            airborne attack threat within 30 minutes Commun.                                                                                                                                       Hot Ring Down (HRD) Telephone System                                 X                  methods Emergency Diesel Generator Buildings                      1, 2, 3, 4, 5          HS1.1           1        2        3          4     5          6      DEF      HA1.1          1       2        3          4     5        6      DEF       HU1.1          1       2        3          4      5      6        DEF                                                                                                                                                                                                                                                        OR Shift Manager Cell Phones                                            X        X Refueling Water Tank                                      1, 2, 3, 4, 5                                                                                                                                                                                                                                                                                                                                                                                                                                                                                               Loss of all Table C-5 NRC communication methods 1                           Water Treatment Plant/Storage Tanks Containment None 1, 2, 3, 4 4, 5 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by PSL Security Shift Supervision A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by PSL Security Shift Supervision A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by PSL Security Shift Supervision OR                                                                                                                                                                                                                                                CA6 Hazardous event affecting SAFETY SYSTEMS needed for the Security                                                                                                                                                                                          OR                                                                       Notification of a credible security threat directed at the site                                                                                                                                                                                        current operating mode A validated notification from NRC of an aircraft attack                     OR                                                                                                                                                                                                                                                CA6.1                                             5       6 threat within 30 min. of the site                                          A validated notification from the NRC providing information of                                                                                                                                Table C-6     Hazardous Events NOTES                                                                                                                                                                    an aircraft threat                                                                                                                                                                                            None                                 The occurrence of any Table C-6 hazardous event Note 1:     The EC should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.                                                                                        [Refer to CA6.1 or SA9.1 for potential escalation                HU2 Seismic event greater than OBE levels                                                6                                                                                              Seismic event (earthquake)
Internal or external FLOODING event AND Event damage has caused indications of degraded 2
performance on one train of a SAFETY SYSTEM needed due to a seismic event]                                                                                                                    Hazardous Note 2:     If an ongoing release is detected and the release start time is unknown,                                                                                                                                                                                                                                                                          None                                               High winds or tornado strike                               for the current operating mode                                                                             None assume   that the release duration has exceeded the specified time limit.                                                                                                                                    HU2.1          1       2       3           4      5      6        DEF                   Event                                                                                                                                                        AND EITHER:
None                                                            None                                                                                              None                                                                                                                          Affecting                                                                                        FIRE Seismic                                                                                                                                                                                                                                                                   Seismic event > OBE as indicated by S-44 CNTMT SEISMIC                                                                                                                                                                                                  Event damage has caused indications of degraded Event                                           Note 3:     If the effluent flow past an effluent monitor is known to have stopped due to                                                                                                                                                                                                                        Safety                                                                                          EXPLOSION MONITOR HIGH LIMIT alarm                                                                                                                                                                                                                                performance to the second train of the SAFETY actions to isolate the release path, then the effluent monitor reading is no                                                                                                                                                                                                                        Systems                                                                                          Other events with similar hazard characteristics                SYSTEM needed for the current operating mode longer VALID for classification purposes.                                                                                                                                                                                                                                                                                                                                                            as determined by the Shift Manager                              Event damage has resulted in VISIBLE DAMAGE to HU3 Hazardous event                                                                                                                                                                                                                                     the second train of the SAFETY SYSTEM needed for
[Refer to CA6.1 or SA9.1 for potential escalation Note 4:     The pre-calculated effluent monitor values presented in EALs RA1.1, due to a hazardous event]                                                                                                                                                                                                                                                                                     the current operating mode RS1.1 and RG1.1 should be used for emergency classification HU3.1          1       2       3           4      5      6        DEF                                                                                                                                                                              (Notes 9, 10) assessments until the results from a dose assessment using actual meteorology are available.                                                                                                                                                                                   A tornado strike within the PROTECTED AREA Note 5:     If the equipment in the listed room or area was already inoperable or out-HU3.2          1       2       3           4      5      6        DEF of-service before the event occurred, then no emergency classification is warranted.                                                                                                                                                                                                   Internal room or area FLOODING of a magnitude sufficient to 3
require manual or automatic electrical isolation of a SAFETY Note 6:     If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-                                                                                                                                          SYSTEM component required by technical specifications for minute None time limit, declaration of a General Emergency is not required None                                                                                              None                                   the current operating mode Natural or Tech.                                             Note 7:     This EAL does not apply to routine traffic impediments such as fog, snow,                                                                                                                                    HU3.3          1       2       3           4      5      6        DEF Hazard                                                         ice, or vehicle breakdowns or accidents.
Movement of personnel within the PROTECTED AREA is Note 8:     A manual trip action is any operator action, or set of actions, which causes                                                                                                                                 IMPEDED due to an event external to the PROTECTED the control rods to be rapidly inserted into the core, and does not include                                                                                                                                  AREA involving hazardous materials (e.g., an offsite chemical manually driving in control rods or implementation of boron injection                                                                                                                                        spill or toxic gas release) strategies.
Note 9:     If the affected SAFETY SYSTEM train was already inoperable or out of                                                                                                                                         HU3.4          1       2       3           4      5      6        DEF service before the hazardous event occurred, then emergency                                                                                                                                                  A hazardous event that results in on-site conditions sufficient classification is not warranted.                                                                                                                                                                             to prohibit the plant staff from accessing the site via personal vehicles (Note 7)
Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY                                                                                                                                           HU4 FIRE potentially degrading the level of safety of the plant SYSTEM, then this emergency classification is not warranted.                                                                             [Refer to CA6.1 or SA9.1 for potential escalation due to fire]
Note 11: One full train of containment heat removal systems consist of any one of                                                                                                                                        HU4.1          1       2       3           4      5      6        DEF the following:                                                                                                                                                                                                  A FIRE is not extinguished within 15 min. of any of the Two (2) Containment Spray Headers with flow in each header at least                                                                                                                                     following fire detection indications (Note 1):
Table H-1 PSL Fire Areas 2,700 gpm                                                                                                                                                                                                    Report from the field (i.e., visual observation)
Four (4) Containment Coolers                                                                                                                                                                                 Receipt of multiple (more than 1) fire zone alarms or Reactor Containment Building/Shield Building                     indications Two (2) Containment Coolers with at least one (1) Containment Spray                                                                         Control Room                                                     Field verification of a single fire zone alarm Header with at least 2,700 gpm                                                                                                              Reactor Auxiliary Building (RAB)                             AND Diesel Generator Building Note 12: This threshold is only applicable if all RVLMS Sensors 4 through 8 are not                                                                          Component Cooling Water (CCW)
The FIRE is located in any Table H-1 area covered inclusive.                                                                                                                                  Intake Structure                                            HU4.2          1       2        3          4      5      6        DEF Intake Cooling Water (ICW)
Note 13: If an RCS heat removal system is in operation within the applicable Table 4                                                                                                                                                                                                                                                                     Receipt of a single fire zone alarm (i.e., no other indications of Steam Trestle Area H                       Fire C-4  heat-up duration and RCS temperature is being reduced, the EAL None not applicable.
None is Note 14: A containment fire zone alarm is considered VALID upon receipt of multiple Yard Areas o RWT o CST a FIRE)
AND Hazards                                                                                                                                                                                                                               o DOST                                                      The fire zone alarm is indicating a FIRE within any Table H-1 zones (more than one) actuated on CP-376 panel.                                                                                                                                                                 area (excluding containment) (Note 14)
Note 15: The Intake Structure includes the Intake Cooling Waiter (ICW) system and                                                                                                                                          AND those immediately surrounding areas that could potentially affect                                                                                                                                               The existence of a FIRE is not verified (proved or disproved) operability or access to ICW piping and components needed for safe                                                                                                                                              within 30 min. of alarm receipt (Note 1) shutdown.                                                                                                                                                                                                       HU4.3          1       2       3           4      5      6        DEF A FIRE within the PROTECTED AREA not extinguished within Table H-2       Safe Operation & Shutdown Rooms/Areas                                                                                                              60 min. of the initial report, alarm or indication (Note 1)
HU4.4          1       2       3           4      5      6        DEF Room/Area                                  Mode A FIRE within the PROTECTED AREA that requires Reactor Auxiliary Building (0.5 and 19.5 levels)           1, 2, 3, 4, 5 firefighting support by an offsite fire response agency to Turbine Building (east end levels required for access to   1, 2, 3, 4, 5                                                                                          extinguish AFW, Steam Trestle, Switchgear and Control Rooms)
Cable Spreading Room                                       1, 2, 3, 4, 5               HA5 Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown A3/B3 Switchgear Rooms                                     1, 2, 3, 4, 5 Steam Trestle Area                                         1, 2, 3, 4, 5               HA5.1          1        2       3          4      5 5                                                          None Intake Structure (Note 15)
Component Cooling Water Area None 1, 2, 3, 4, 5 1, 2, 3, 4, 5 Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area                                                                                None Hazardous                                                                                                  Emergency Diesel Generator Buildings                       1, 2, 3, 4, 5                 AND Gases                                                                                                    Refueling Water Tank                                       1, 2, 3, 4, 5 Entry into the room or area is prohibited or IMPEDED (Note 5)
Water Treatment Plant/Storage Tanks                         1, 2, 3, 4 Containment                                                     4, 5 HS6 Inability to control a key safety function from outside the Control       HA6 Control Room evacuation resulting in transfer of plant control to Room                                                                          alternate locations HS6.1            1       2        3          4     5          6               HA6.1          1       2        3          4     5        6      DEF 6
An event has resulted in the Control Room being evacuated                     An event has resulted in the Control Room being evacuated None AND                                                                                                                                                                                        None Control of any of the following key safety functions is not Control re-established within 15 min. of the last licensed operator Room leaving the Control Room (Note 1):
Evacuation Reactivity (modes 1, 2 and 3 only)
Core cooling RCS heat removal HG7 Other conditions exist which in the judgment of the EC warrant                        HS7 Other conditions existing that in the judgment of the EC warrant          HA7 Other conditions exist that in the judgment of the EC warrant          HU7 Other conditions existing that in the judgment of the EC warrant declaration of a General Emergency                                                        declaration of a Site Area Emergency                                          declaration of an Alert                                                    declaration of an Unusual Event HG7.1               1       2        3          4         5      6            DEF       HS7.1            1       2        3          4     5          6      DEF      HA7.1          1       2        3          4     5        6      DEF       HU7.1          1       2        3          4      5      6        DEF Other conditions exist which, in the judgment of the EC,                                Other conditions exist which, in the judgment of the EC,                      Other conditions exist which, in the judgment of the EC,                   Other conditions exist which, in the judgment of the EC, 7                    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 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 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 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 EC                                                                                                              results in intentional damage or malicious acts, (1) toward site HOSTILE ACTION that results in an actual loss of physical                                                                                                             probable life threatening risk to site personnel or damage to              been initiated. No releases of radioactive material requiring Judgment control of the facility. Releases can be reasonably expected                            personnel or equipment that could lead to the likely failure of               site equipment because of HOSTILE ACTION. Any                              offsite response or monitoring are expected unless further to exceed EPA Protective Action Guideline exposure levels                               or, (2) that prevent effective access to equipment needed for                 releases are expected to be limited to small fractions of the              degradation of SAFETY SYSTEMS occurs.
offsite for more than the immediate site area.                                          the protection of the public. Any releases are not expected                   EPA Protective Action Guideline exposure levels.
to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.
EU1 Damage to a loaded cask CONFINEMENT BOUNDARY EU1.1        1      2      3       4     5      6     DEF E                        1                                                             None                                                                                 None                                                                        None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HSM) > any of the following:
Confinement ISFSI               Boundary                                                                                                                                                                                                                                                                         1,600 mrem/hr on the HSM front bird screen 10 mrem/hr on the outside HSM door centerline 10 mrem/hr on the HSM end shield wall exterior St. Lucie Plant Modes:                                                         1 Power Operation 2
Startup                Hot Standby 3                        4 Hot Shutdown             Cold Shutdown 5                      6 Refueling DEF Defueled UNIT 2 Emergency Action Level Matrix                                                              EAL WALL CHART - COLD CONDITIONS (RCS  200&deg;F)
Revision X Prepared for FPL by Operations Support Services, Inc. 4-19-19}}