SBK-L-17131, Radiological Emergency Plan (Ssrep), Revisions 72, 73 and Er 1.1, Classification of Emergencies, Revision 58

Radiological Emergency Plan (Ssrep), Revisions 72, 73 and Er 1.1, Classification of Emergencies, Revision 58

ML17221A228
Person / Time
Site:	Seabrook
Issue date:	08/09/2017
From:	Browne K NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
SBK-L-17131
Download: ML17221A228 (200)

Text

NEXTera ENERGY ~

SEABROOK August 9, 2017 Docket No. 50-443 Docket No. 72-63 SBK-L-17131 United States Nuclear Regulatory Commission Attn.: Document Control Desk Washington, D.C. 20555-0001 Seabrook Station Radiological Emergency Plan (SSREP), Revisions 72, 73 and ER 1.1, Classification of Emergencies, Revision 58 In accordance with the requirements of 10 CFR 50, Appendix E; 10 CFR 50.4, and 10 CFR 72.44(f), enclosed are Revisions 72 and 73 to the Seabrook Station Radiological Emergency Plan (SSREP) and Revision 58 to ER 1.1, Classification of Emergencies.

The revisions do not reduce the effectiveness of the SSREP, and the SSREP continues to meet the standards of 10 CFR 50.47(b) and 10 CFR 50, Appendix E. The Resident Inspector copy is provided directly through the NextEra Energy Seabrook, LLC records management system. provides a summary of changes to the SSREP and ER 1.1. Enclosure 2 provides a summary of the change analysis required by 10 CFR 50.54(q)(5) and Enclosure 3 provides a copy of the revised manual and procedure.

Should you have any questions regarding the enclosed revisions, please contact me at (603) 773-7932.

Sincerely, NextEra Energy Seabrook, LLC PO Box 300 , Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L-17131/Page2 cc (with enclosures):

J.P. DeBoer, Region I, Division of Reactor Safety cc (without Enclosure 3):

ATTN: Document Control Desk Director, Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards Nuclear Regulatory Commission Washington, DC 20555 -0001 J. Poole, NRC Project Manager, Project Directorate 1-2 cc (without enclosures):

D . H. Dorman, NRC Region I Administrator P. Cataldo, NRC Senior Resident Inspector

Enclosure 1 to SBK-L-17131 Summary of Changes Radiological Emergency Plan (SSREP), Revision 72

• Section 5 - Revised EAL description to match new NRC approved EAL scheme.

Editorial change to replace referenced procedure from the NARC to LI-AA-102-IOOI with regard to regulatory reporting.

Radiological Emergency Plan (SSREP), Revision 73

• Section 5 - Corrected footers on EAL charts. Corrected typographical error on Figure 5.7 (Category A to Category R)

• Section I2 - Replaced validation, exemption, and deferral are discussed in the ERO Training Program Description with training requirements are discussed in the ERO Training Program Description.

• Appendix D- Updated Letter of Agreement with the State of New Hampshire and Commonwealth of Massachusetts.

ER 1.1, Classification of Emergencies, Revision 58

• Complete rewrite to new EAL scheme based on NEI 99-0I rev. 06 per LAR I5-02 (Amendment I 52). The new revision of ER I. I merges the emergency classification procedure with the Emergency Action Level Design Basis Document. The following changes were made to the basis document after the safety evaluation was received:

o HUI -- Corrected the reference to possible upgrade path from HGI to HG7.

o HU2, MA9, CA6 -- EALs describe events 'as determined by the Shift Manager."

This should be STED/SED to be consistent with the remainder of the documents.

o MAI - Removed incorrect statement in basis that was copied over from MS I.

o A diagram was added to the fission product barrier section to assist the STED/SED when making a judgement call regarding a loss of the fuel cladding.

Enclosure I, Page I of I

En.closure 2 to SBK-L-17131 Change Analysis Summary Radiological Emergency Plan (SSREP), Revision 72 Revision 72 updated the description of new Emergency Action Level (EAL) scheme based on NEI 99-01, Revision 6. The new EAL scheme was submitted for NRC preapproval under License Amendment Request 15-02, and approved by NRC safety evaluation ML16358A411 finding that, "the Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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 public."

The 10 CFR 50.54(q) evaluation determined the IOCFR50.47(b)(4), function and timing of the function of maintaining a standard scheme of emergency classification and action levels are maintained because the site continues to use an NRC approved EAL scheme

[Amendment 152].

An editorial change replaced referenced procedure NARC to LI-AA-102-1001 since the information regarding regulatory reporting is now contained in the LI-AA- I 02-1001 procedure. This change is not associated with a planning standard.

Radiological Emergency Plan (SSREP), Revision 73 Revision 73 made editorial corrections to Section 5, replaced reference to specific elements of the training program being discussed within the ERO Training Program Description (TPD) with a generic reference to the TPD, and updated the Letter of Agreement between the site and the State of New Hampshire and Commonwealth of Massachusetts.

With regard to the TPD reference revision, the 10 CFR 50.54(q) evaluation determined the 10CFR50.47(b)(15) function and timing of the function of providing training to emergency responders is maintained because there is no decrease in the training provided to ERO members. The fleet no longer allows required training to be deferred which is more conservative since an ERO member must now either complete or follow the fleet exemption process from training by the required due date. The timing of the function is maintained because there is no change to the annual training requirements referenced in the SSREP or change in training frequency.

With regard to updating the Letter of Agreement, the 10 CFR 50.54(q) evaluation determined the 10CFR50.47(b)(3) function of making arrangements for requesting and using offsite assistance is maintained because an active letter of agreement between the NextEra, the State of New Hampshire, and Commonwealth of Massachusetts, is preserved. There is no timing associated with this function.

Enclosure 2, Page 1 of 3

ER 1.1, Classification of Emergencies, Rev. 58:

The procedure was rewritten to adopt the new EAL scheme based on NEI 99-01 rev. 06 per LAR 15-02. The new revision of ER 1.1 merges the emergency classification procedure with the Emergency Action Level Design Basis Document, providing the user a single document to refer to when classifying emergencies. The new EAL scheme was submitted for NRC preapproval under License Amendment Request 15-02, and approved by NRC safety evaluation ML16358A411 finding that, "the Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that 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 public."

The 10 CFR 50.54(q) evaluation determined the 10CFR50.47(b)(4), function and timing of the function of maintaining a standard scheme of emergency classification and action levels are maintained because the site continues to use an NRC approved EAL scheme

[Amendment 152].

The following changes were made to the basis document after the safety evaluation was received:

HUl -- Corrected the reference to possible upgrade path from HGl to HG7.

The original EAL submittal to the NRC included EAL HG 1, which declared a general emergency based on a hostile action security event. During the NRC review, the NRC issued EPFAQ 2015-13 (ML16166A366), which concluded that licensees may exclude HG 1 from their EAL basis provided that EALs RA2, RS2, RG2, RSl, RGl, HSl , HS6, HS7, and HG7 are as endorsed and, therefore, bound the events of concern for EAL HG 1.

The EAL scheme was revised to eliminate HG 1, however the potential upgrade paths listed in HU4 continued to list HG 1 as a potential upgrade path and should have replaced HG 1 with HG7. HG7 contains the latitude for the STED/SED to declare a discretionary general emergency for events including those associated with a hostile action security event.

The NRC concluded in the safety evaluation (ML16358A411) that excluding HG 1 was acceptable.

The 10 CFR 50.54(q) evaluation determined the 10CFR50.45(b)(4) function of maintaining a standard scheme of emergency classification and action levels is maintained because the correct reference is now included in the basis.

HG 1 no longer exists.

Providing the correct upgrade reference does not adversely affect the timing of the function of maintaining a standard scheme of Providing the correct upgrade reference do emergency classification and action levels.

HU2, MA9, CA6 -- EALs describe events 'as determined by the Shift Manager." This should be STED/SED to be consistent with the remainder of the documents . HU2, MA9, CA6 allow a classification to be made based on judgment. While the standard NEI 99-01 rev. 06 template language refers to the Shift Manager making the determination, the intent is that the judgment of Enclosure 2, Page 2 of 3

EAL applicability rests with the individual responsible for making the classification. While the responsible individual is likely the Shift Manager, it is more appropriate to list both the Short Term Emergency Director and Site Emergency Director since either could be responsible depending on the emergency conditions. STED/SED determination is also consistent with the remainder of the document.

The 10 CFR 50.54(q) evaluation determined the 10CFR50.47(b)(4) function of maintaining a standard scheme of emergency classification and action levels is maintained because the intent of who is responsible for making the classification is more accurate, and does not change the intent of the EAL.

The timing of the function is maintained because accurately defining the individual responsible for classification allows for prompt classification and eliminates any confusion should there be any concern that the Shift Manager must personally endorse the determination.

MAl - Removed incorrect statement in basis that was copied over from MS 1.

The general basis description of SEPS functionality is common between MAI and MS 1. MS 1 is a total loss of all AC power and the basis contains a statement that MS 1 is not applicable if a vital bus is being powered from SEPS, which is correct. This statement was inadvertently included in the MAI basis, which is not correct.

The 10 CFR 50.54(q) evaluation determined the 10CFR50.47(b)(4) function of maintaining a standard scheme of emergency classification and action levels is maintained because the incorrect information was removed from the basis which could have prevented the EAL from correctly being classified.

There is no timing associated with this function.

A diagram was added to the fission product barrier section to assist the STED/SED when making a judgement call regarding a loss of the fuel cladding. The previous EAL Design Basis Document contained a figure that assists the STED/SED when attempting to make a discretionary judgment regarding the loss of the fuel cladding barrier on the fission product barrier matrix. Since the diagram continues to provide accurate information, it was included at the request of the Operators.

The 10 CFR 50.54(q) evaluation determined the 10CFR50.47(b)(4) function of maintaining a standard scheme of emergency classification and action levels is maintained because the diagram assists the operators in making a discretionary decision regarding the loss of the fuel cladding barrier and does not alter the intent of the EAL.

There is no timing associated with this function.

The above changes do not reduce the effectiveness of the emergency plan and the emergency plan as changed continues to meet the planning standards of 10 CFR 50.4 7 (b) and the requirements of 10 CFR 50, Appendix E.

Enclosure 2, Page 3 of 3 to SBK-L-17131 CHANGE INSTRUCTIONS SEABROOK STATION RADIOLOGICAL EMERGENCY PLAN (SSREP)

Page 1of1 REMOVE 1 _ INSERT Cover Cover Rev. 73 Table of Contents Table of Contents Rev. 73 List of Effective Pages List of Effective Pages Rev. 73 Chapter 1: .Chapter 1:

1-5.1thru1-5.12 *1-5.1thru1-5.12 Rev. 71 Figu,res 5.6 thru 5.8 Figures 5.6 thru 5.8 Rev. 57 1-12.1thru1-12.8 1-12.1thru1-12.8 Rev. 71 1-13.1 1-13.1 and 1-13.2 Rev. 72 Appendix D Appendix D Cover Page Cover Page Rev. 62 Table Of Contents Table of Contents Rev. 62 D-5 thru D-5c D-5 thru D-5c Rev. 57 D-10 D-10 Rev. 62

RMD Controlled Copy _ _ __

SEABROOK STATION PROGRAM MANUAL Seabrook Station Radiological Emergency Plan SS REP Manual Owner:

Rev. 73 D. Currier

SEABROOK STATION RADIOLOGICAL EMERGENCY PLAN (SS REP)

TABLE OF CONTENTS CONTENT PAGE

1.0 INTRODUCTION

1-1.1 2.0 DEFINITIONS 1-2.1 3.0 RADIOLOGICAL EMERGENCY PLAN

SUMMARY

1-3.1 3.1 Introduction 1-3.1 3.2 Station Emergency Response 1-3.1 3.3 Local and State Government Responses 1-3.3 3.4 Federal Government Response 1-3.3 Figure 3.1 Notification Plan Figure 3.2 Relationship of the Seabrook Station ERO to Offsite Organizations 4.0 THE AREA 1-4. l 4.1 The Site 1-4.1 4.2 Area Characteristics, Land Use and Demography 1-4. l 4.2.1 Area Characteristics 1-4.1 4.2.2 Uses of Adjacent Lands and Waters 1-4.2 4.2.3 Population Distribution 1-4.2 4.3 Emergency Planning Zones 1-4.2 Table 4.1 Summary of Peak Population Estimates of Communities within 0 to 10 Miles of the Site Table 4.4 Communities Within the Seabrook Station Plume Exposure Pathway Emergency Planning Zone Figure 4.1 Site Boundaries Figure 4.2 Major Routes in 10 Mile Study Figure 4.3 Site Layout Figure 4.4 2010 Resident Population Distribution within a 0-10 Mile Radius of Seabrook Station Figure 4.6 Estimate Peak Transient Population (0-10 Miles)

Figure 4.7 Seabrook Station "Plume Exposure" Emergency Planning Zone Figure 4.8 Seabrook Station "Ingestion Exposure" Emergency Planning Zone (County Designations)

Page 1 SSREP Rev. 73

CONTENT PAGE 5.0 EMERGENCY CLASSIFICATION SYSTEM 1-5.l 5.1 Regulatory Context 1-5.1 5.2 Definitions Used in Developing EAL Methodology 1-5.2 5.3 Recognition Categories 1-5.3 5.4 Emergency Class Descriptions 1-5.5 5.5 Emergency Class Thresholds 1-5.7 5.6 Emergency Action Levels 1-5.8 5.7 Treatment of Multiple Events and Emergency Class Upgrading 1-5.9 5.8 Emergency Class Downgrading 1-5.10 5.9 Classifying Transient Events 1-5.10 5.10 Cold Shutdown/Refueling IC/EALs 1-5.11 5.11 ISFSI IC/EALs 1-5 .11 Figure 5.6 Emergency Initiating Condition Matrix - Modes 1, 2, 3 and 4 Figure 5.7 Emergency Initiating Condition Matrix - Modes 5, 6 and Defueled Figure 5.8 Fission Product Barrier Degradation Matrix - Modes 1, 2, 3 and 4 6.0 EMERGENCY FACILITIES AND EQUIPMENT 1-6.1 6.1 Emergency Centers 1-6.1 6.1.1 Technical Support Center 1-6.1 6.1.2 Operational Support Center 1-6.1 6.1.3 Emergency Operations Facility 1-6.2 6.1.4 Support for Radiological Analysis of Environmental Samples 1-6.3 6.1.5 Joint Information Center 1-6.3 6.1.6 Federal Radiological Monitoring and Assessment Center 1-6.3 6.2 Assessment Capability 1-6.4 6.2. 1 Process Monitors 1-6.4 6.2.2 Radiation Data Management System 1-6.4 6.2.3 Geophysical Phenomena Monitors 1-6.5 6.2.4 Fire Detection Systems 1-6.6 6.2.5 Facilities and Equipment for Offsite Monitoring 1-6.6 Figure 6.1 Location of Emergency Operation Centers Around the Seabrook Station Site Page 2 SSREP Rev. 73

CONTENT Figure 6.2 Relative Location of Technical Support within the 75' Elevation Level of the Control Building Figure 6.5 Operational Support Center Layout Figure 6.6 EOF Layout 7.0 COMMUNICATIONS 1-7.1 7.1 Nuclear Alert System 1-7.l 7.2 NRC Communications Channels 1-7.1 7.3 Telephone System 1-7.2 7.4 Commercial Pager Service 1-7.2 7.5 Station Radio System 1-7.2 7.5.1 Offsite Monitoring Team Radio Network 1-7.2 7.5 .2 UHF Radio System 1-7.3 7.6 Station Paging System 1-7.4 7.7 Sound-Powered Telephone System 1-7.4 Figure 7.1 Emergency Notification Figure 7.2 Coordination Channels with States Figure 7.3 Offsite Monitoring Team Radio Communications Figure 7.4 Telephone Communication Systems Overview Figure 7.5 UHF Radio Communication Systems Overview 8.0 ORGANIZATION 1-8.1 8.1 Introduction 1-8.1 8.2 Emergency Response Organization 1-8.1 8.2.1 On-Shift Emergency Response Organization 1-8.1 8.2.2 Augmented Emergency Response Organization 1-8.2 8.3 Emergency Public Information Organization 1-8.3 8.4 Seabrook Station Corporate Support 1-8.3 8.5 Recovery Organization 1-8.4 8.6 Extensions of Seabrook Station Emergency Response Organization 1-8.4 8.6.1 Local Services 1-8.4 8.6.2 Federal Government Support 1-8.5 8.6.3 Private Organization Support 1-8.5 Page 3 SSREP Rev. 73

CONTENT PAGE 8.7 Coordination with State Government Authorities 1-8.5 Figure 8.1 On-Shift Emergency Response Organization Figure 8.2 Augmented Emergency Response Organization for Unusual Event Figure 8.3 Augmented Emergency Response Organization for Alert, Site Area Emergency, and General Emergency Figure 8.4 Emergency Operations Facility Staff Figure 8.5 Operational Support Center Staff Figure 8.6 Technical Support Center (TSC) Staff Figure 8.7 Canceled Figure 8.9 Joint Information Center Staff Figure 8.12 On-Shift Emergency Response Organization Actions Figure 8.13 Summary of the Radiological Emergency Responsibilities and Functions of the Massachusetts State Authorities Figure 8.14 Summary of the Radiological Emergency Responsibilities and Functions of the New Hampshire State Authorities Figure 8.15 Comparison of NUREG-0654 Emergency Response Staffing Goals with the Seabrook Station Emergency Response Organization (ERO)

Figure 8.16 Seabrook Station News Services Staff 9.0 EMERGENCY RESPONSE OUTLINE 1-9.1 9.1 Initiation 1-9.1 9.2 Activation of the Emergency Organization 1-9. l 9.2.1 Unusual Event Response 1-9.1 9.2.2 Alert Response 1-9.2 9.2.3 Site Area Emergency Response 1-9.4 9.2.4 General Emergency Response 1-9.5 9.3 Emergency De-escalation, Termination and Recovery 1-9.5 Figure 9.1 Method of Notification and Reporting Instructions for Onsite Personnel 10.0 EMERGENCY MEASURES 1-10.l 10.1 Radiological Accident Assessment Systems and Techniques 1-10.1 10 .1.1 Estimation of Offsite Dose Rates 1-10.2 10.1.2 Evaluation of Field Environmental Samples 1-10.3 10.1.3 Evaluation of Post Accident Samples 1-10.4 10 .1.4 Severe Accident Management Guidance 1-10.4 10.2 Protective Action Recommendation Criteria 1-10.4 10.3 Radiological Exposure Control 1-10.5 Page 4 SSREP Rev. 73

CONTENT PAGE 10.4 Protective Measures 1-10.6 10.4.1 Personnel Accountability 1-10.6 10.4.2 Station Access/Egress Control Methods 1-10.6 10.4.3 Protective Measures for Hostile Action Based Events 1-10.7 10.4.4 Decontamination Capability 1-10.8 10.4.5 Use of Onsite Protective Equipment and Supplies 1-10.8 10.4.6 Radiation Guideline Action Levels 1-10.8 10.5 Aid to Affected Personnel 1-10.9 10.5.1 Medical Treatment 1-10.9 10.5 .2 Medical Transportation 1-10.9 Table 10.1 EPA Protective Action Guidelines 1-10.10 Table 10.2 Emergency Dose Limits 1-10.11 Table 10.3 Emergency Center Protection 1-10.12 Figure 10.1 Emergency Center Protection Figure 10.2 Seabrook Station Evacuation Routes 11.0 EMERGENCY NOTIFICATION AND PUBLIC INFORMATION 1-11.1 11.1 Emergency Notification 1-11.1 11 .2 Public Notification 1-11.1 11.3 Public Information 1-11.1 12.0 MAINTAINING EMERGENCY PREPAREDNESS 1-12.1 12.1 Drills and Exercises 1-12.1 12.1.1 Radiological Emergency Plan Exercises 1-12.1 12.1.2 Emergency Plan Drills 1-1 2. 1 12.1.3 Drill and Exercise Scenarios 1-12.3 12.1.4 Evaluation of Exercises 1-12.4 12.1.5 Credit for Response to an Actual Emergency 1-12.4 12.2 Emergency Plan Training 1-12.5 12.2.1 Emergency Response Organization (ERO) 1-12.5 12.2.2 Support Groups 1-12.6 12.2.3 Station Personnel with No ERO Assignment 1-12.6 12.2.4 Emergency Preparedness Department Personnel 1-1 2.6 12.2.5 Records 1-12.6 12.3 Review and Updating of Plan and Procedures 1-12.7 12.4 Maintenance and Inventory of Emergency Equipment and Supplies 1-12.7 Page 5 SSREP Rev. 73

CONTENT PAGE 12.5 Emergency Preparedness Manager 1-12.7 12.6 Technical Training Supervisor 1-12.8 12.7 Operations Support Manager 1-12.8 13.0

SUMMARY

OF CHANGES 1-13.l APPENDICES Appendix A Emergency Response Organization Position Definitions A-1 Appendix B Canceled B-1 Appendix C Evacuation Time Estimates C-1 Appendix D Letters of Agreement with Emergency Response Organizations D-1 Appendix E Seabrook Station Public Alert and Notification System E-1 Appendix F Emergency Equipment F-1 Appendix G Seabrook Station Supporting Emergency Plans and Procedures Listing G-1 Appendix H NUREG-0654/Seabrook Station Radiological Emergency Plan Cross Reference H-1 Page 6 SSREP Rev. 73

LIST OF EFFECTIVE PAGES PAGE REV. PAGE REV.

Cover 73 Figure 8.5 59 Figure 8.6 58 TOC 1 - 6 73 Figure 8.7 Canceled Figure 8.9 64 LOEP 1 -2 73 Figure 8.12 50 1-1.1and1-1.2 56 Figure 8.13 13 Figure 8.14 52 1-2.1 thru 1-2.5 66 Figure 8.15 Sheet 1 55 1-3.1 thru 1-3.4 66 Sheet 2 55 Figure 3.1 64 Sheet 3 55 Figure 3.2 55 Sheet 4 55 Sheet 5 55 1-4.1 thru 1-4.5 66 Sheet 6 55 Figure 4.1 56 Figure 8.16 32 Figure 4.2 20 Figure 4.3 58 1-9 .1 thru 1-9. 7 66 Figure 4.4 66 Figure 9.1 66 Figure 4.6 66 Figure 4.7 5 1-10.1thru1-10.12 70 Figure 4.8 Undated Figure 10.1 30 Figure 10.2 48 1-5.1 thru 1-5 .12 71 Figure 5.6 57 1-11.1thru1-11.3 64 Figure 5.7 57 1-12.1thru1-12.8 71 Figure 5.8 57 1-13.1 thru 1-13 .2 72 1-6.1thru1-6.7 70 Figure 6.1 56 Appendix A Figure 6.2 42 Cover Page 64 Figure 6.5 47 Index 64 Figure 6.6 64 A-1 thru A-25 64 1-7. 1 thru 1-7.4 64 Appendix B cancellation sheet 42 Figure 7.1 64 Figure 7.2 64 Appendix C Figure 7.3 64 Cover Page 49 Figure 7.4 64 C-1 thru C-5 49 Figure 7.5 64 Appendix D 1-8.1 thru 1-8.6 70 Cover Page 62 Figure 8.1 70 Table of Contents 62 Figure 8.2 60 D-1 58 Figure 8.3 60 D-la 58 Figure 8.4 66 D-lb 58 D-2 thru D-21 61 Page 1 SSREP Rev. 73

LIST OF EFFECTIVE PAGES PAGE REV. PAGE D-3 thru D-3b 46 D-4 61 D-5 thru D-5c 57 D-6 61 D-7 59 D-8 thru D-8e 59 D-9 thru D-9b 60 D-10 62 Appendix E Cover Page 57 E-1 thru E-5 57 AppendixF 33 Appendix G G-1 thru G-7 55 AppendixH Cover Page 47 H-1 thru H-7 47 Page 2 SSREP Rev. 73

5.0 EMERGENCY CLASSIFICATION SYSTEM Seabrook Station uses NEI 99-01, Revision 6, as the basis for the emergency classification system. The information in this chapter is derived from generic basis discussion presented in NEI 99-01, Revision 6.

5.1 Regulatory Context Title 10, Code of Federal Regulations, Part 50 provides the regulations that govern emergency preparedness at nuclear power plants. Nuclear power reactor licensees are required to have NRC-approved "emergency response plans" for dealing with "radiological emergencies." The requirements call for both onsite and offsite emergency response plans, with the offsite plans being those approved by FEMA and used by the State and local authorities. This section deals with the utilities' approved onsite plans and procedures for response to radiological emergencies at nuclear power plants, and the links they provide to the offsite plans.

Section 50.47 of Title 10 of the Code ofFederal Regulations (10 CFR 50.47), entitled "Emergency Plans," states the requirement for such plans. Part (a)(l) of this regulation states that "no operating license will be issued unless a finding is made by NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency."

The major portion of 10 CFR 50.47 lists "standards" that emergency response plans must meet.

The standards constitute a detailed list of items to be addressed in the plans. Of particular importance to this project is the fourth standard, which addresses "emergency classification" and "action levels." These terms, however, are not defined in the regulation.

10 CFR 50.54, "Conditions of licenses," emphasizes that power reactor licensees must "follow, and maintain in effect, emergency plans which meet the standards in Part 50.47(b) and the requirements in Appendix E to this part." The remainder of this part deals primarily with required implementation dates.

10 CFR 50.54(q) allows licensees to make changes to emergency plans without prior Commission approval only if: (a) the changes do not decrease the effectiveness of the plans and (b) the plans, as changed, continue to meet 10 CFR 50.47(b) standards and 10 CFR 50 Appendix E requirements. The licensee must keep a record of any such changes. Proposed changes that decrease the effectiveness of the approved emergency plans may not be implemented without application to and approval by the Commission.

10 CFR 50.72 deals with "Immediate notification requirements for operating nuclear power reactors." The "immediate" notification section actually includes three types ofreports:

(1) immediately after notification of State or local agencies (for emergency classification events);

(2) one-hour reports; and, (3) four-hour reports.

Although 10 CFR 50.72 contains significant detail, it does not define either "Emergency Class" or "Emergency Action Level." But one-hour and four-hour reports are listed as "non-emergency events," namely, those which are "not reported as a declaration of an Emergency Class." Certain 10 CFR 50.72 events can also meet the Unusual event emergency classification if they are precursors of more serious events. These situations also warrant anticipatory notification of state and local officials. (See Section 3.7, "Emergency Class Descriptions".)

1-5.1 SSREP Rev. 71

By footnote, the reader is directed from 10 CFR 50.72 to 10 CFR 50 Appendix E, for information concerning "Emergency Classes."

10 CFR 50.73 describes the "Licensee event report system," which requires submittal of follow-up written reports within sixty days of required notification ofNRC.

10 CFR 50 Appendix E, Section B, "Assessment Actions," mandates that emergency plans must contain "emergency action levels." EALs are to be described for: (1) determining the need for notification and participation of various agencies, and (2) determining when and what type of protective measures should be considered. Appendix E continues by stating that the EALs are to be based on: (1) in-plant conditions; (2) in-plant instrumentation; (3) onsite monitoring; and (4) offsite monitoring.

10 CFR 50 Appendix E, Section C, "Activation of Emergency Organization," also addresses "emergency classes" and "emergency action levels." This section states that EALs are to be based on: (1) onsite radiation monitoring information; (2) offsite radiation monitoring information; and, (3) readings from a number of plant sensors that indicate a potential emergency, such as containment pressure and the response of the Emergency Core Cooling System. This section also states that "emergency classes" shall include: (1) Unusual events (UNUSUAL EVENTs), (2) Alert, (3) Site Area Emergency, and (4) General Emergency.

These regulations are supplemented by various regulatory guidance documents. A significant document that has dealt specifically with EALs is NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," October 1980.

5.2 Definitions Used in Developing EAL Methodology The following definitions apply to the Seabrook Station EAL methodology:

EMERGENCY CLASS: One of a minimum set of names or titles, established by the Nuclear Regulatory Commission (NRC), for grouping off-normal nuclear power plant conditions according to (1) their relative radiological seriousness, and (2) the time-sensitive onsite and off-site radiological emergency preparedness actions necessary to respond to such conditions.

The existing radiological emergency classes, in ascending order of seriousness, are called:

• Unusual event

• Alert

• Site Area Emergency

• General Emergency INITIATING CONDITION (IC): One of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency, or such an emergency has occurred.

1-5.2 SSREP Rev. 71

Discussion:

In NUREG-0654, the NRC introduced, but does not define, the term "initiating condition." Since the term is commonly used in nuclear power plant emergency planning, the definition above has been developed and combines both regulatory intent and the greatest degree of common usage among utilities.

Defined in this manner, an IC is an emergency condition which sets it apart from the broad class of conditions that may or may not have the potential to escalate into a radiological emergency. It can be a continuous, measurable function that is outside technical specifications, such as elevated RCS temperature or falling reactor coolant level (a symptom). It also encompasses occurrences such as FIRE (an event) or reactor coolant pipe failure (an event or a barrier breach).

EMERGENCY ACTION LEVEL (EAL): A pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given emergency class. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (on site or offsite); a discrete, observable event; results of analyses ; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency class.

Discussion:

The term "emergency action level" has been defined by example in the regulations, as noted in the above discussion concerning regulatory background. The term had not, however, been defined operationally in a manner to address all contingencies. There are times when an EAL will be a threshold point on a measurable continuous function, such as a primary system coolant leak that has exceeded technical specifications for a specific plant.

5.3 Recognition Categories ICs and EALs can be grouped in one of several schemes. This generic classification scheme incorporates symptom-based, event-based, and barrier-based ICs and EALs.

The symptom-based category for ICs and EALs refers to those indicators that are measurable over some continuous spectrum, such as core temperature, coolant levels, containment pressure, etc. When one or more of these indicators begin to show off-normal readings, reactor operators are trained to identify the probable causes and potential consequences of these "symptoms" and take corrective action. The level of seriousness indicated by these symptoms depends on the degree to which they have exceeded technical specifications, the other symptoms or events that are occurring contemporaneously, and the capability of the licensed operators to gain control and bring the indicator back to safe levels.

Event-based EALs and ICs refer to occurrences with potential safety significance, such as the failure of a high-pressure safety injection pump, a safety valve failure, or a loss of electric power to some part of the plant. The range of seriousness of these "events" is dependent on the location, number of contemporaneous events, remaining plant safety margin, etc.

1-5.3 SSREP Rev. 71

Barrier-based EALs and ICs refer to the level of challenge to principal barriers used to assure containment ofradioactive materials contained within a nuclear power plant. For radioactive materials that are contained within the reactor core, these barriers are: fuel cladding, reactor coolant system pressure boundary, and containment. The level of challenge to these barriers encompasses the extent of damage (loss or potential loss) and the number of barriers concurrently under challenge. In reality, barrier-based EALs are a subset of symptom-based EALs that deal with symptoms indicating fission product barrier challenges. These barrier-based EALs are primarily derived from Emergency Operating Procedure (EOP) Critical Safety Function (CSF)

Status Tree Monitoring (or their equivalent). Challenge to one or more barriers generally is initially identified through instrument readings and periodic sampling. Under present barrier-based EALs, deterioration of the reactor coolant system pressure boundary or the fuel clad barrier usually indicates an "Alert" condition, two barriers under challenge a Site Area Emergency, and loss of two barriers with the third barrier under challenge is a General Emergency. The fission product barrier matrix described in Category F is a hybrid approach that recognizes that some events may represent a challenge to more than one barrier, and that the containment barrier is weighted less than the reactor coolant system pressure boundary and the fuel clad barriers.

Symptom-based ICs and EALs are most easily identified when the plant is in a normal startup, operating or hot shutdown mode of operation, with all of the barriers in place and the plant's instrumentation and emergency safeguards features fully operational as required by technical specifications. It is under these circumstances that the operations staff has the most direct information of the plant's systems, displayed in the main control room. As the plant moves through the decay heat removal process toward cold shutdown and refueling, barriers to fission products are reduced (i.e. , reactor coolant system pressure boundary may be open) and fewer of the safety systems required for power operation are required to be fully operational. Under these plant operating modes, the identification of an IC in the plant's operating and safety systems becomes more event-based, as the instrumentation to detect symptoms of a developing problem may not be fully effective; and engineered safeguards systems, such as the Emergency Core Cooling System (ECCS), are partially disabled as permitted by the plant's Technical Specifications.

Barrier-based ICs and EALs also are heavily dependent on the ability to monitor instruments that indicate the condition of plant operating and safety systems. Fuel cladding integrity and reactor coolant levels can be monitored through several indicators when the plant is in a normal operating mode, but this capability is much more limited when the plant is in a refueling mode, when many of these indicators are disconnected or off-scale. The need for this instrumentation is lessened, however, and alternate instrumentation is placed in service when the plant is shut down.

It is important to note that in some operating modes there may not be definitive and unambiguous indicators of containment integrity available to control room personnel. For this reason, barrier-based EALs should not place undue reliance on assessments of containment integrity in all operating modes. Generally, Technical Specifications relax maintaining containment integrity requirements in modes 5 and 6 in order to provide flexibility in performance of specific tasks during shutdown conditions. Containment pressure and temperature indications may not increase ifthere is a pre-existing breach of containment integrity. At most plants, a large portion of the containment's exterior cannot be monitored for leakage by radiation monitors.

1-5.4 SSREP Rev. 71

Several categories of emergencies have no instrumentation to indicate a developing problem, or the event may be identified before any other indications are recognized. A reactor coolant pipe could break; FIRE alarms could sound; radioactive materials could be released; and any number of other events can occur that would place the plant in an emergency condition with little warning. For emergencies related to the reactor system and safety systems, the ICs shift to an event based scheme as the plant mode moves toward cold shutdown and refueling modes. For non-radiological events, such as FIRE, external floods, wind loads, etc., as described in NUREG-0654 Appendix 1, event-based ICs are the norm.

In many cases, a combination of symptom-, event- and barrier-based ICs will be present as an emergency develops. In a loss of coolant accident (LOCA), for example:

• Coolant level is dropping; (symptom)

• There is a leak of some magnitude in the system (pipe break, safety valve stuck open) that exceeds plant capabilities to make up the loss; (barrier breach or event)

• Core (coolant) temperature is rising; (symptom) and

• At some level, fuel failure begins with indicators such as high coolant activity samples, etc. (barrier breach or symptom) 5.4 Emergency Class Descriptions There are three considerations related to emergency classes. These are:

(1) The potential impact on radiological safety, either as now known or as can be reasonably projected; (2) How far the plant is beyond its predefined design, safety, and operating envelopes; and (3) Whether or not conditions that threaten health are expected to be confined to within the site boundary.

The ICs deal explicitly with radiological safety impact by escalating from levels corresponding to releases within regulatory limits to releases beyond EPA Protective Action Guideline (PAG) plume exposure levels. In addition, the "Discussion" sections below include offsite dose consequence considerations which were not included in NUREG-0654 Appendix 1.

UNUSUAL EVENT: Events are in process 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.

1-5.5 SSREP Rev. 71

Discussion:

Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant technical specification Limiting Condition of Operation (LCO) allowable action statement time for achieving required mode change. Precursors of more serious events should also be included because precursors do represent a potential degradation in the level of safety of the plant. Minor releases of radioactive materials are included. In this emergency class, however, releases do not require monitoring or offsite response.

ALERT: Events are in process 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.

Discussion:

Rather than discussing the distinguishing features of "potential degradation" and "potential substantial degradation," a comparative approach would be to determine whether increased monitoring of plant functions is warranted at the Alert level as a result of safety system degradation. This addresses the operations staffs need for help, independent of whether an actual decrease in plant safety is determined. This increased monitoring can then be used to better determine the actual plant safety state, whether escalation to a higher emergency class is warranted, or whether de-escalation or termination of the emergency class declaration is warranted. Dose consequences from these events are small fractions of the EPA PAG plume exposure levels, i.e., about 10 mrem to 100 mrem TEDE.

SITE AREA EMERGENCY: Events are in process or have occurred which involve an actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Discussion:

The discriminator (threshold) between Site Area Emergency and General Emergency is whether or not the EPA PAG plume exposure levels are expected to be exceeded outside the site boundary. This threshold, in addition to dynamic dose assessment considerations discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response agency concerns as to timely declaration of a General Emergency.

GENERAL EMERGENCY: Events are in process 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.

1-5.6 SSREP Rev. 71

Discussion:

The bottom line for the General Emergency is whether evacuation or sheltering of the general public is indicated based on EPA PAGs, and therefore should be interpreted to include radionuclide release regardless of cause. To better assure timely notification, EALs in this category must primarily be expressed in terms of plant function status, with secondary reliance on dose projection. In terms of fission product barriers, loss of two barriers with loss or potential loss of the third barrier constitutes a General Emergency.

5.5 Emergency Class Thresholds The most common bases for establishing these boundaries are the technical specifications and setpoints for each plant that have been developed in the design basis calculations and the Updated Final Safety Analysis Report (UFSAR).

For those conditions that are easily measurable and instrumented, the boundary is likely to be the EAL (observable by plant staff, instrument reading, alarm setpoint, etc.) that indicates entry into a particular emergency class. For example, the main steam line radiation monitor may detect high radiation that triggers an alarm. That radiation level also may be the setpoint that closes the main steam isolation valves (MSIV) and initiates the reactor trip. This same radiation level threshold, depending on plant-specific parameters, also may be the appropriate EAL for a direct entry into an emergency class.

In addition to the continuously measurable indicators, such as coolant temperature, coolant levels, leak rates, containment pressure, etc., the UFSAR provides indications of the consequences associated with design basis events. Examples would include steam pipe breaks, MSIV malfunctions, and other anticipated events that, upon occurrence, place the plant immediately into an emergency class.

Another approach for defining these boundaries is the use of a plant-specific probabilistic safety assessment (PSA - also known as probabilistic risk assessment, PRA). A PSA has been completed for Seabrook Station. PSAs can be used as a good first approximation of the relevant ICs and risk associated with emergency conditions for existing plants. Generic insights from PSAs and related severe accident assessments which apply to EALs and emergency class determinations are:

1. Prolonged loss of all AC power events are extremely important. This would indicate that should this occur, and AC power is not restored within 15 minutes, entry into the emergency class at no lower than a Site Area Emergency, when the plant was initially at power, would be appropriate. This implies that precursors to loss of all AC power events should appropriately be included in the EAL structure.

2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting containment integrity may be difficult in these conditions. This is why maintaining containment integrity alone following sequences leading to severe core damage may be an insufficient basis for not escalating to a General Emergency.

1-5.7 SSREP Rev. 71

3. EAL methodology must be sufficiently rigorous to cover risk-significant sequences such as containment bypass, large LOCA with early containment failure, station blackout greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (e.g., LOCA consequences of Station Blackout), and reactor coolant pump seal failure.

Another critical element of the analysis to arrive at these threshold (boundary) conditions is the time that the plant might stay in that condition before moving to a higher emergency class. In particular, station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 1.155, "Station Blackout," is used to determine whether a Site Area Emergency or a General Emergency is indicated. The time dimension is critical to the EAL since the purpose of the emergency class for state and local officials is to notify them of the level of mobilization that may be necessary to handle the emergency. This is particularly true when a "Site Area Emergency" or "General Emergency" is imminent.

Regardless of whether or not containment integrity is challenged, it is possible for significant radioactive inventory within containment to result in EPA PAG plume exposure levels being exceeded even assuming containment is within technical specification allowable leakage rates.

With or without containment challenge, however, a major release of radioactivity requiring offsite protection actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents,"

indicates that such conditions do not exist when the amount of clad damage is less than 20%.

5.6 Emergency Action Levels With the emergency classes defined, the thresholds that must be met for each EAL to be placed under the emergency class can be determined. There are two basic approaches to determining these EALs. EALs and emergency class boundaries coincide for those continuously measurable, instrumented ICs, such as radioactivity, core temperature, coolant levels, etc. For these ICs, the EAL will be the threshold reading that most closely corresponds to the emergency class description using the best available information.

For discrete (discontinuous) events, the approach will have to be somewhat different. Typically, in this category are internal and external hazards such as fire or earthquake. The purpose for including hazards in EALs is to assure that station personnel and offsite emergency response organizations are prepared to deal with consequential damage these hazards may cause. If, indeed, hazards have caused damage to safety functions or fission product barriers, this should be confirmed by symptoms or by observation of such failures. Therefore, it may be appropriate to enter an emergency for events approaching or exceeding design basis limits such as Operating Basis Earthquake, design basis wind loads, FIRE within VITAL AREAs, etc. This would give the operating staff additional support and improved ability to determine the extent of plant damage. If damage to barriers or challenges to Critical Safety Functions (CSFs) have occurred or are identified, then the additional support can be used to escalate or terminate the Emergency Class based on what has been found. Of course, security events must reflect potential for increasing security threat levels.

1-5.8 SSREP Rev. 71

Plant emergency operating procedures (EOPs) are designed to maintain and/or restore a set of CSFs which are listed in the order of priority for restoration efforts during accident conditions.

The Seabrook Station CSF set includes:

• Subcriticality

• Core cooling

• Heat sink

• Pressure-temperature-stress (RCS integrity)

• Containment

• RCS inventory

• Emergency Coolant Recirculation

• Radiation/RDMS Display There are diverse and redundant plant systems to support each CSF. By monitoring the CSFs instead of the individual system component status, the impact of multiple events is inherently addressed, e.g., the number of operable components available to maintain the critical safety function.

The EOPs contain detailed instructions regarding the monitoring of these functions and provides a scheme for classifying the significance of the challenge to the functions . In providing EALs based on these schemes, the emergency classification can flow from the EOP assessment rather than being based on a separate EAL assessment. This is desirable as it reduces ambiguity and reduces the time necessary to classify the event.

As an example, consider that the Westinghouse Owner's Group (WOG) Emergency Response Guidelines (ERGs) classify challenges as YELLOW, ORANGE, and RED paths. If the core exit thermocouples exceed 1, 100 degrees F or 725 degrees F with low reactor vessel water level, a RED path condition exists. The ERG considers a RED path as 11

... an extreme challenge to a plant function necessary for the protection of the public ... This is almost identical to the present 11 NRC NUREG-0654 description of a site area emergency 11

... actual or likely failures of plant functions needed for the protection of the public ... It reasonably follows that if any CSF enters a 11 RED path, a site area emergency exists. A general emergency could be considered to exist if core cooling CSF is in a RED path and the EOP function restoration procedures have not been successful in restoring core cooling.

5.7 Treatment of Multiple Events and Emergency Class Upgrading The emergency class declared is based on the highest EAL reached. For example, two Alerts remain in the Alert category. Or, an Alert and a Site Area Emergency is a Site Area Emergency.

1-5 .9 SSREP Rev. 71

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

5.8 Emergency Class Downgrading Another important aspect of usable EAL guidance is the consideration of what to do when the risk posed by an emergency is clearly decreasing. Seabrook Station uses a combination approach involving recovery from General Emergencies and some Site Area Emergencies and termination from Unusual Events, Alerts, and certain Site Area Emergencies causing no long-term plant damage. Downgrading to lower emergency classes adds notifications but may have merit under certain circumstances.

5.9 Classifying Transient Events For some events, the condition may be corrected before a declaration has been made. For example, an emergency classification is warranted when automatic and manual actions taken within the control room do not result in a required reactor trip. However, it is likely that actions taken outside of the control room will be successful, probably before the STED/SED classifies the event. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses (e.g., coolant radiochemistry sampling, may be necessary).

If the emergency-related indications completely clear before a declaration of an emergency classification level has been made, then no emergency classification is required. The Shift Manager shall notify the Emergency News Manager within one hour of the termination of the emergency-related indications that emergency-related indications briefly existed, but cleared prior to the declaration of an emergency classification. The Emergency News Manager will initiate state notifications per good neighbor notification procedures. The event shall be reported to the NRC in accordance with 10 CFR 50.72 and 50.73 per LI-AA-102-1001, Regulatory Reporting, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the event.

If emergency-related indications are received and later cleared, and after the fact it is determined that an emergency classification was warranted but not made, then no emergency classification is required. The Shift Manager shall notify the Emergency News Manager within one hour of discovery that an emergency classification was warranted but not declared and that emergency-related indications no longer exist. The Emergency News Manager will initiate state notifications per good neighbor notification procedures. The event shall be reported to the NRC in accordance with 10 CFR 50.72 and 50.73 per LI-AA-102-1001, Regulatory Reporting, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the event.

If emergency-related indications are received and reduce in severity, such that the emergency classification went from an earlier higher level to a current lower level, the current lower level emergency should be declared.

1-5.10 SSREP Rev. 71

Reporting requirements of 10 CFR 50. 72 are applicable and the guidance of NUREG-1022, Rev. 1, Section 3 should be applied.

5.10 Cold Shutdown/Refueling IC/EALs Generic Letter 88-17, Loss of Decay Heat Removal, SECY-91-283, Evaluation of Shutdown and Low Power Risk Issues, SECY-93-190, Regulatory Approach to Shutdown and Low-power Operation, 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, all address nuclear power plant safety issues that are applicable to periods when the plant is shutdown. These evaluations identify a number of variables which significantly affect the probability and consequences of losing decay heat removal capability during shutdown periods. In addition, NUREG-1449 discusses that the need to respond appropriately, including emergency classification and notification, still exists during cold-shutdown and refueling conditions. Through use ofNEI 99-01, Revision 6, the Seabrook Station emergency classification system addresses issues concerning shutdown effects on declaring emergencies discussed in SECY-93-190 and NUREG-1449.

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 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 modes.

The initiating conditions and example emergency actions levels associated directly with Cold Shutdown or Refueling safety function are presented in Recognition Category C, Cold Shutdown/Refueling.

5.11 ISFSI IC/EALs 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. The Final Rule governing Emergency Planning Licensing Requirements for Independent Spent Fuel Storage Facilities (Federal Register Volume 60, Number 120 June 22, 1995, Pages 32430-32442) indicated that a significant amount of the radioactive material contained within a cask must escape its packaging and enter the biosphere for there to be a significant environmental impact resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.

1-5.11 SSREP Rev. 71

Recognition Category E (Events Related to ISFSI) is applicable to licensees using their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32. The emergency classifications for Recognition Category E are those provided by NUREG 0654/FEMA Rep.1 in accordance with 10 CFR 50.47. The classification of an ISFSI event under provisions of a 10 CFR 50.47 emergency plan should be consistent with the definitions of the emergency classes as used by that plan.

1-5.12 SSREP Rev. 71

EMERGENCY INITIATING CONDITION MATRIX Modes 1, 2, 3, and 4 ALERT Category R- Abnonnal Rad Levels/Radiological Effluent

~~1~1~i~~ f~ ~~~~~U:o~~d~o~~~;~Y mrem T EDE or 5 ,000 mrem thyroid i

~

Ij ~~1~1~i~~ ~~ ~~~~~U:o~~d~~~~vi~rem TEDE or 500 mrem thyroid COE i RA 1 ! ~~i~~~~i~~y9~e5seu~t~~g0rnli~~~~te dose i > 10 mrem TE DE or 50 mrem thyroid Release of gaseous or liquid radioactivity> 2 times the OOCM limits for~ 60 minutes

-~;~1>1_~~~-s:~/I.. .. . . .. ... . . ** * ***** h-. l ~~. ::~~.s . ~fJ.. . . .. . . .... . . . . . . . . . . . . .L. . . . . l.~;~1>1_~~~-s:.~/I. . . . .. . . . . . ... . . .. ..

Op. Modes: All

~:S~~::~~~~~~~~5~e~.~a~.n~Cebv~l J) ~ - i ~e~f~el p:~; level at 1.5 n. (Level 3)  ! RA2 ! ;b~~:~~~~l~~:~~~ 0~fi~=~~~t~~~~el. UNPLANNED loss of water level above irradiated fuel for 60 minutes or longer. r j p. o es:  !  ! Op. Modes: All Op. Modes: All

.9.f'.:_!:!_~<!-~~;*_12~-----------------------------..;. .......... L----------------------------------------------------~---********t*****************--********------****

~ [ RA3  ! ~a~~~\i~~~en~e~!~~=~~~f~~~o~~~~ss i f  ! plant operations, shutdown or f f j cooldown.

:  ! Op. Modes: All Category E - Events Related to ISFSI Malfunction Damage to a loaded cask CONFINEMENT BOUNDARY Op. Mode: All Category H - Hazards and Other Conditions Affe_~!!!.'.~..~!~!.'.!_~~!.~!Y...._ ... -*******-**-**-*-, - - - , - - - - - - - - - - - - -

1~~~i~~T~~T~~~.ilhin the HA1  ! ~~~T~~ECAiJ~~~~~~; lt;:EA or Confirmed SECURITY CONDIT ION or threat.

j Op. Modes: All l airborne attack threat within 30 Op. Modes: All t .... .... . ... . . .. ..... . .. .J J.~~".;:d~E:~~IJ. . . Seismic event greater than OBE I levels .

..':JP-..¥.~~!:.~:~~~/

I Hazardous event.

Op. Modes: All FIRE potentially degrading the level of safety of the plant.

1.

. I

.9e--~~?.~~.:~~~----------------------------

i... f .HAiT*G-aseous release impeding access to 1  ! equipment necessary for normal

'......l_i * -*-*-* *-* *-*-*-*- -*-*- - -*- -*- *- -*- - -*- 1! * -* *

• l! _~;~~~~;;~:;s_~:~~:~ ~ -~:............

Inability to control a key safety function HAS Control Room evacuation resulting in i from outside the Control Room  !  ! transfer of plant control to alternate j Op. Modes: All  ! i locations

.L---------------------------------------------------~--------L9-P.:_~_~C!_~~.:~~~------------------------------

Other conditions exist which in the judgment of the STED/SED warrant j ~~hge~~~~~~t~~~ssef~b~h~~ ~na~~:nt [ HA7 i ~~hge~~~~~~t~~~sse;~b~h~~ :a~~:nt Other conditions exist which in the judgment of the STED/SED warrant declaratio n of a General Emergency j declaration of a Site Area Emergency  ! i declaration of an Alert declaration of an Unusual Event Op. Modes: Afl  ! Op. Modes: All f i Op. Modes: All Op. Modes: Alf Category M - System Malfunction Prolonged loss of all offsite and all l Loss of all offsite and all onsite AC  ! MA1 Loss of all but one AC power source Loss of all offsite AC power on site AC power to emergency  ! power to emergency buses for 15  !':,

to emergency buses for 15 minutes capability to emergency buses for buses j minutes or longer or longer 15 minutes or longer AND  ! Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 Restoration of at least one AC emergency bus in less than 4 I I l

hours is not likely.

OR  !

Core Cooling (C) CSF RED entry conditions met j  !

_<?J?.:_~~~!'..~.:-_!.!_?:_~:_+/- ____________________ _ ***ow***l *--*-----------------------------------------------1---*-***+-***-*****************-*-**-****-*****--*-**--

i  ! MA2 ! UNPLANNED loss of Control Room UNPLANNED loss of Control Room l 1  ! indications for 15 minutes or longer indications for 15 minutes or longer

!  !  ! with a significant transient in Op. Modes: 1, 2, 3, 4 I!

l_ _ _ __J__~;~~~~~-~:.!~l~-~!-~------------------*-*

. Reactor coolant activity greater than Technical Specification allowable j i limits Op. Modes: 1, 2, 3, 4 ll i :

RCS leakage for 15 minutes or tonger Op. Modes: 1, 2, 3, 4

........1---------------------------------------------------l--------t---*-*******-***--*--*--**********-****-***********

i Inability to shutdown the reactor  ; MAS ! Automatic or manual trip fail s to Automatic or manual trip fails to l causing a challenge to core cooling or *  ! shutdown the reactor, and shutdown the reactor j RCS heat removal  ! subsequent manual actions taken at Op. Modes: 1

• Op. Modes: 1  ! the Main Control Board are not

! successful in shutting down the

!  :! reactor

---

*-t* ----------L-<:!E.*__ ~9.~!:~~---!

r********1 Loss of all onsite or offsite communicatio ns capabilities

[ j Op. Modes: 1, 2, 3, 4 Failure to isolate containment or loss of containment pressure control Loss of all AC and Vital DC power

! --i_~;~-~f~ii"\iii'.;i"jj"C"P~~;~-f~-~---------------- !

i Op Modes: 1, 2, 3, 4

- ~'-'~"-ur.: .:; .:d;:.;:; .o:;1_1"-~-=;'-i';"-~-'-~*-s_o_ri_on_g_*_r--+--1--~~m~~;;;,_o;,I~~~~~ ......J...........

! MA9 l.. ... . ....

! Hazardous event affecting a

' i SAFETY SYSTE M needed for the i current

! Op. Modes: 1, 2, 3, 4

_ _ - - - - -~~~~~ _1! f.! ~ -~~<! ~------.

Figure 5.6 SSREP Rev. 57

EMERGENCY INITIATI NG CONDITION MATRIX Modes 5, 6, and Defu eled GENERAL EMERGENCY SITE AREAEMER ALERT Category R - Abnormal Rad Levels/Radiological Effl uent Release of gaseous radioactivity i Release of gaseous radioactivity j RA1 j Release of gaseous or liquid Release of gaseous or liquid resulting in offsite dose > 1,000 j resulting in offs1te dose> 100 mrem *.': 1 radioactivity resulting in offsite dose radioactivity> 2 times the ODCM mrem TEDE or 5,000 mrem thyroid ~ TEDE or 500 mrem thyroid COE j > 10 mrem TEDE or 50 mrem thyroid limits for~ 60 minutes

-~~~~~".o~--~~------------------------------1.-------'j -~:-~-~~~~--~~---------------------------------l------J~~~~-~".o~:~~~-- - ----------- ----------------

Op. Modes: All spent fuel pool level cannot be - Spent fuel pool level at 1.5 ft. (Level 3)  ! RA2 i Significant lowering of water level UNPLANNED loss of water level restored to at least 1.5 ft. (Level 3) i Op. Modes: All  ! [ above, or damage to, irradiated fuel . above irradiated fuel for 60 minutes or longer. j  :  : Op. Modes: All Op. Modes: All

.9P.:.M!:!.~~~:~~~------------------------------ ********t---------------------------------------------------L-J*-**************--****-***-------------..-----------------

[ l  ! RAJ I~a:~~\i~~~~~e~~~~:~~~f~~;o~~~~ss i j  ! j plant operations, shutdown or r i  ! j ~;l:;::s: All Category E - Eve nts Related to ISFSI Malfunction Damage to a loaded cask CONFINEMENT BOUNDARY Op. Mode: All Category H - Hazards and Other Conditio ns Affecting Plant Safety i HOSTILE ACTION within the HA1 HOSTILE ACTION within the Confirmed SECURITY CONDITION i PROTECTED AREA . OWNER CONTROLLED AREA or or threat j Op. Modes: All  ! ai~borne attack threat within 30 Op. Modes: All r*-*r-- ------- - ------ ---

1

! minutes

, , , .... "' -- --- - - Seismic event greater th an OBE levels.

Op. Modes: All Hazardous event.

Op. Modes: All i l i F IR E potentially degrading the level ii  !

l i

f i of safety of the plant.

.9-EJ.:.M..?.<!..:~.:~~~-----------*----------------

: HAS r**Gaseous release impeding access to i  !  ! equipment necessary for normal

: i plant operations, shutdown or

~  !  ! cooldown

~

f-rj l

• - *1~-;;-biiilY"t~-~~;t;~j-~-k.;y*;~f~iY"f~-~~ti~-~-- ! "HAG-- j

! _O~p_.M_o_d_e_s:_A_ll_ _ _ _ _ __

Control Room evacuation resulting in f from outside the Control Room  !  ! transfer of plant control to alternate

! 1= Op. Modes: All  !  ! locations

---

*---*****----*******T**

. i i Op. Modes: All

• ot'****************************************--------*-*T********-t--------------------------------------------------

Qther conditions exist which in the  ! j Other conditions exist which in the l HA7 j Other conditions exist which in the Other conditions exist which in the judgment of the STED/SED warrant  ! j judgment of the STED/SED warrant 1 l judgment of the STED/SED warrant judgment of the STED/SED warrant declaration of a General Emergency  ! j declaration of a Site Area Emergency  ! j declaration of an Alert declaration of an Unusual Event Op. Modes: All f  ! Op. Modes: All  ! j Op. Modes: All Op. Modes: All Category C - Cold Shutdown/Refu eling System Malfunction Loss of reactor vessel/RCS inventory j Loss of reactor vessel/RCS inventory CA1  ! Loss of reactor vessel/RCS inventory UNPLANNED loss of reactor affecting fuel clad integrity with containment challenged

! affecting core decay heat removal j capability I Op. Modes: 5, 6 vessel/RCS inventory for 15 minutes or longer Op. Modes: 5, 6 I  ! Op. Modes: 5, 6

. . .i. c-.A.iT.Lo;~-~f-an I 1,.. Op. Modes: 5, 6

• r-**-*****-*-r ------------------------- offsile and all onsite A Loss of all but one AC power

power to emergency buses for 15 source to emergency buses for 15 l I i minutes or longer minutes or longer I ,ri **c-.i\3 9E.*--~~~-:'.~:-..~:..?.~_Defueled Inability to maintain the plant in cold shutdown Op. Modes: 5, 6, Defue/ed UNPLANNED increase in RCS temperature.

Op. Modes: 5, 6 OR Loss of ALL RCS temperature and reactor vessel/RCS level indication J for 15 minutes or longer i

• -*-****-*~****************** Of?.:_Modes: 5, 6 ----*--*-***********---------------

Loss of Vital DC power for 15 minutes or longer Op. Modes: 5, 6 Loss of all onsite or offsite communicatio ns capabilities

!..........[__________________________________________________________ Op. Modes: 5, 6, Defueled

! GAS i ~!~~~u~~~~~~~:~~;d afor the j current operating mode

• • -**--**-~ 9E.*-~~~::-~~*--~!__?. ________ _

- -- -- - -- - -- -- ----- - ---- - ----~

~ ___ ~-~d_e:_s_ ~~ ~*-'!~~ _l'!_e_f~~!e_d____ ~

Figure 5.7 SSREP Rev . 57

FISSION PRODUCT BARRIER DEGRADATION MATRIX Modes 1, 2, 3, and 4 Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier Sub-Category Loss Potential Loss Loss Potent ial Loss Los s Potential L oss RCS Integrity (P) RED entry conditions met Core Cooling (C) RED entry conditions met for Core Cooling (C) ORANGE entry conditions 15 minutes or longer met with RCS press> 300 psig.

Core Cooling (C) RED entry conditions met. OR

1. CSF Status OR OR Containment (Z) CSF - RED entry condiiions (Note 1)

Heat Sink (H) RED entry conditions met. Heat Sink (H) RED entry conditions met. met.

(Note 1) (Note 1) (Note 1)

RCS activity> 300 uCi/gm Dose Equivalent 1-131 (as determined per Procedure

2. RCS Activity CS0925.01, Reactor Coolant Post Accident Samolinal Operation of a second charging pump in the An automatic or manual SI actuation is

3. RCS Leakage required by EITHER of the following:

1. UNISOLABLE RCS leakage normal charging mode is required by EITHER of the following:

1. UNISOLABLE RCS leakage IIndications of RCS leakage outside of containment.

OR OR

2. SG tube RUPTURE

2. SG tube leakage.

A leaking or RUPTURED SG is FAULTED

4. S/G Rupture or Faull outside of containment.

Containment isolation is required Containment H2 concent ration ~ 6%

AND EITHER of the following: OR

1. Containment integrity has been lost 1. Containment pressure > 18 psig

5. Containment Integrity based on STED/ SEO judgment. AND OR 2. Less than one full train of Containment

2. UNISOLABLE pathway from the Building Spray (CBS) is operating per containment to the environment exists. desian for 15 minutes or lonaer

6. Containment Post-LOCA Radiation Monitors Post-LOCA Radiation Monitors Post-LOCA Radiation Monitors Radiation Monitor RM-6576A-1 or RM-65768-1 ~ 95 R/hr RM-6576A- 1 or RM-6576B-1 ~ 16 R/ hr RM-6576A-1 or RM-65768-1 ~ 1,305 R/hr

7. STE D/S ED Judgment A ny condition in the opinion of the STEO/SED A ny condition in the o~inion of the STED/SED that indicates a Potential Loss of the Fuel Clad IAny condition in the opinion of the ST.ED/SEO that indicates a Loss of the RCS Barner.

Any condition in the opinion of the STED/SED that indicates a Potential Loss of the RCS Any condition in the opinion of the STED/SED that indicates a Loss of the Containment Any condition in the opinion of the STED/SED that indicates a Potential Loss of the that indicates a Loss of the Fuel Clad Barrier.

Barrier. Barrier. Barrier. Containment Barrier.

Barrier St at us Alert FA1 - ANY Loss or Potential Loss of EITHER Fuel Clad or RCS Barriers Fuel Clad Loss Enter

.I -+

Fuel Clad Polential Loss Enter

.I -+

RCS Loss Enter

.I -+

RCS Potential Loss Enter

.I -+

Containment Loss Enter

.I -+

Containment Potential Loss Enter

.I -+

Emergency Class1ficat1on ~ m:I I s:rn:I I S'l'ra r s:rn:I l r:JlTa I lffi1:I I s:rn:I r m:I I ~ r r:JlTa I lffi1:I I r:JlTa I ALERT I ALE RT I ALERT I ALERT NOTE 1: Refer to ER 1.1, Section 1.1, Discussion concerning the proper use of CSFSTs as EALs Figure 5.8 SSREP Rev. 57

12.0 MAINTAINING EMERGENCY PREPAREDNESS 12.1 Drills and Exercises Emergency exercises and drills shall be conducted to test and evaluate the adequacy of emergency facilities, equipment, procedures, communication channels, actions of emergency response personnel, and coordination between Seabrook Station and offsite agencies. A summary of exercises and drills, and associated elements are presented below.

As used for emergency preparedness drills and exercises, "annual" means that the event shall be conducted once within a calendar year. For "semi-annual," the event shall be conducted once within the first 6 calendar months of a year and once again within the second 6 calendar months.

"Biennial" means the event will be conducted within a two-year period.

12.1.1 Radiological Emergency Plan Exercises An exercise tests the execution of the overall Station emergency response and its integration with responding offsite organizations. In order to test and evaluate the Station emergency response, an exercise shall be conducted every two years. Consistent with the regulatory requirements for offsite exercise participation, Federal, State and local agencies shall be notified of intended exercises and their conduct shall be coordinated with offsite authorities as appropriate.

12.1.2 Emergency Plan Drills A drill is a supervised instruction period aimed at testing, developing and maintaining skills in a particular emergency response function. The frequency of drills is dependent upon the function to be tested.

1. Combined Functional Drills To ensure that adequate emergency response capabilities are maintained during the interval between biennial exercises, at least one annual drill will be conducted involving a combination of some of the principal functional areas of the onsite emergency response capabilities. The principal functional areas of emergency response include activities such as management and coordination of emergency response, accident assessment, protective action decision making, and plant system repair and corrective actions. Activation of all of the emergency response facilities will not be necessary during these drills. State and local governments within the plume exposure pathway EPZ may participate in these drills at their request.

2. Communication Drills To ensure that emergency communications equipment is operable, communication drills shall be conducted as outlined below. Included in the scope of these drills is the aspect of understanding message content. Paragraphs c, d, and g below may be performed as part of annual combined functional drills and the required biennial exercise.

1-12.1 SSREP Rev. 71

a. Communication channels with State governments within the plume exposure pathway shall be tested monthly;

b. The pager system for the notification of the Primary Responders of the Emergency Response Organization (ERO) shall be tested weekly;

c. Data transmission capability between Station emergency centers shall be tested annually;

d. EOF communications to State Emergency Operation Centers and to Station field assessment teams shall be conducted annually;

e. Communications between the Control Room and the NRC Headquarters Operations Center shall be tested weekly or as otherwise directed by the NRC;

f. Communications between the EOF, TSC and the NRC Headquarters Operations Center shall be tested monthly or as otherwise directed by the NRC; and

g. Notification of the Secondary Responders of the ERO via the RapidNotify emergency notification service shall be tested at least annually.

3. Fire Drills To evaluate the response and training of the Station fire brigade and coordination of same with offsite fire support, a number of fire drills are conducted annually with at least one drill being conducted with offsite fire support. The drills shall be conducted in accordance with the Seabrook Station Fire Protection Manual (SSFP).

4. Medical Drills To evaluate the response and training of the Station medical response and offsite hospital personnel, a medical drill shall be conducted annually involving a simulated contaminated individual. Although the Station medical response may be tested more frequently, the offsite response portion of medical drills may be performed as part of the biennial exercise.

5. Radiological Monitoring Drills Plant environs and radiological monitoring drills (onsite and offsite) shall be conducted annually. These drills shall include collection and analysis of airborne sample media, communications, recordkeeping and, if feasible, interface with other offsite monitoring efforts. In addition, a drill will be conducted on the collection of other sample media (e.g., soil, water and vegetation). Radiological monitoring drills may be performed as part of a training activity, another drill or the biennial exercise.

1-12.2 SSREP Rev. 71

6. Health Physics Drills Health Physics drills shall be conducted semiannually which involve response to, and analysis of, simulated elevated airborne and liquid samples and direct radiation measurements. These drills may be performed as part of a training activity, another drill or the biennial exercise. Additionally, Chemistry personnel shall be drilled annually on obtaining and analyzing post-accident samples.

12.1.3 Drill and Exercise Scenarios The Emergency Preparedness Manager is responsible for coordinating preparation for and implementation of drills and exercises with the exception of fire and medical emergency drills. Operations Support staff are responsible for coordinating preparation for and implementation of fire and medical emergency drills. For exercises that include offsite participation, the scenario shall be submitted to FEMA for agency review in accordance with regulatory guidance. All exercise scenarios shall be submitted to the NRC prior to implementation.

Within an eight-year period (beginning 1/1/2014), drill and exercise scenario content shall be varied to test all the major elements of the emergency preparedness program.

These major elements correspond to the objectives presented in applicable fleet and site procedures. Within an eight-year period, one scenario shall include the states' response within the ingestion pathway EPZ. In general, the scenario shall simulate a sequence of emergency conditions that would call for the mobilization of the offsite authorities, require recommendations of offsite protective measures, and allow for evaluation of offsite plans and their integration with the Station response. The scenario shall include, as a minimum, the following:

1. Date, time period, locations and participating organizations;

2. Basic objectives and specific elements that are to be tested;

3. Guidelines and extent of play;

4. Controller instructions, and a list of controllers and evaluators;

5. A narrative summary of the exercise scenario and expected responses; and

6. Time schedule of real and simulated events.

Seabrook Station cannot commit other organizations to conduct an exercise during off-hour times. Outside organizations shall be encouraged to participate in exercises, but starting times and pre-notification for exercises have to be agreed upon by participating offsite organizations. Exercises will be conducted in different seasons of the year, to the extent practicable, depending on circumstances such as scheduled refueling outages and exercise schedules for other sites affecting the availability of NRC and FEMA evaluators.

1-12.3 SSREP Rev. 71

The exercise shall be structured with sufficient flexibility to allow free play for decision-making processes. The exercise scenario package shall describe a specific accident sequence, contain a set of input messages, and list anticipated response actions which parallel the accident sequence. The exercise controller organization shall receive instructions to recognize areas where ERO responses may deviate from anticipated responses. The exercise controller organization may (1) restrict player action ifthe response threatens the approved time sequence; (2) restrict player action if the response circumvents a required exercise objective; and (3) introduce "free play" items to the scenario sequence if player actions become stagnant.

Exercise elements which may allow free play in the decision-making process include the following:

1. Exposure control actions;

2. Manpower augmentation actions;

3. Emergency classification actions, particularly the de-escalation process;

4. Recommendation of protective actions; and

5. Coordination and communication with offsite authorities.

12.1.4 Evaluation of Exercises To evaluate the performance of participating facility personnel and the adequacy of emergency facilities, equipment and procedures used during an exercise, the Exercise Manager shall arrange for qualified controllers and evaluators to evaluate and critique the exercise.

A critique shall be conducted as soon as feasible following the conclusion of the exercise with player personnel as designated by the Exercise Manager. After the critique, the controllers and evaluators shall provide drill/exercise-related documentation and performance reports to the Drill/Exercise Manager. The Drill/Exercise Manager shall use this information to determine whether, and to what extent, drill/exercise objectives were demonstrated.

The exercise documentation shall be submitted to the Emergency Preparedness Manager who shall assign responsibility and deadlines for corrective actions. Individuals assigned this responsibility shall be required to document actions taken to improve the Station's emergency preparedness.

12.1.5 Credit for Response to an Actual Emergency Demonstration of exercise or drill objectives scheduled for evaluation in accordance with Fleet EP Drill and Exercise procedures may be satisfied by the effective response and documentation of designated key ERO staff to an actual emergency. Credit will be given for this objective when the following provisions are met in response to an actual emergency.

1-12.4 SSREP Rev. 71

1. The emergency required a prompt and timely response and mobilization of key ERO staff responsible for the implementation of RERP emergency functions;

2. The emergency resulted in the establishment of communications links among responding organizations;

3. The following documentation, describing the level of response and involvement of key ERO staff to the emergency, is available:

Type of emergency; Period of response; Arrival times of responders; Communications logs; Emergency decisions made and implemented; Emergency plan resources used; and List of staff involved .

4. The event is evaluated in accordance with Emergency Preparedness Department procedures for Post Event Reviews and Evaluations to determine if the actions taken were appropriate or the response warrants implementation of future corrective measures.

12.2 Emergency Plan Training The following sections describe the various types of Emergency Plan Training.

12.2.1 Emergency Response Organization (ERO)

Training for ERO personnel is conducted in accordance with the ERO Training Program Description. Changes to this document shall be reviewed to ensure that (1) they do not decrease the effectiveness of the SSREP, the SSER or Seabrook Station emergency response capabilities, and (2) when implemented, the emergency preparedness program will continue to meet the applicable standards of 10 CFR 50.47(b) and the requirements of 10 CFR 50, Appendix E. (Protected: Ref. NRC Inspection Report 50-443/93-03)

Major elements of the program are discussed below.

Seabrook Station personnel with specific positions in the ERO shall receive training to initially qualify them for a response position. ERO assignments shall, as much as possible, parallel normal job knowledge, skills and abilities.

Initial training shall consist of an overview course and other courses that are appropriate to the individual's response position. The required initial courses are specified in the ERO Training Program Description.

1-12.5 SSREP Rev. 71

Selected ERO members shall receive annual re-qualification training to maintain their level of emergency response knowledge. The required re-qualification training courses are also specified in the ERO Training Program Description. Re-qualification training courses are conducted throughout the year. The ERO Training Program Description contains a generic annual schedule which is used to ensure that re-qualification training occurs at about the same time period each year. Re-qualification courses may be scheduled up to three months away from the generic schedule to accommodate plant events such as outages.

Annual re-qualification training courses shall be completed within 15 months. Training requirements are discussed in the ERO Training Program Description.

Training other than that shown in the ERO Emergency Preparedness Training Program Description may be given to address specific needs.

12.2.2 Support Groups Personnel from support groups who report to Seabrook Station shall be offered training designed to aid them in performing their emergency response function, including the Town of Seabrook Fire Department. This training shall be offered annually.

Support groups that do not report to Seabrook Station shall also be offered training designed to aid them in performing their emergency response function. These personnel include NH Homeland Security & Emergency Management, NH Department of Health and Human Services, Massachusetts Emergency Management Agency, Massachusetts Department of Public Health, Maine Emergency Management Agency, Wentworth-Douglass Hospital and Exeter Hospital. (Protected: Ref. NRC IR 86-1 8[03])

This training shall be offered annually.

12.2.3 Station Personnel with No ERO Assignment Station personnel with no ERO assignment shall be trained in their proper response to an emergency during Plant Access Training. This training shall be given on an annual basis.

12.2.4 Emergency Preparedness Department Personnel Emergency Preparedness Department personnel receive plant access training and training specific to their individual ERO assignments. In addition, the Emergency Preparedness Manager schedules personnel participation in specialized emergency planning training, participation in EP related conferences, and as technical specialists for EP audits at other sites.

12.2.5 Records Documentation of training conducted in support of emergency planning is maintained in accordance with appropriate nuclear training procedures.

1-12.6 SSREP Rev. 71

12.3 Review and Updating of Plan and Procedures Independent reviews of the Seabrook Station emergency preparedness program shall be conducted every 12 months. The reviews shall include the emergency plan, its implementing procedures, training, equipment, readiness testing and State and local government planning interfaces. Management controls shall be implemented for evaluation and correction of review findings . The result of the review, along with recommendations for improvements, shall be documented and retained for a period of five years.

Intent revisions to the SSREP and to SSER emergency plan implementing procedures ER 1.1 ,

Classification of Emergencies; ER 1.2, Emergency Plan Activation; and ER 5.4, Protective Action Recommendations, shall be submitted to the Station Operation Review Committee (SORC) for review and approval before implementation. Intent revisions of other emergency plan implementing procedures contained in the SSER shall be reviewed by a station qualified reviewer per the Station Qualified Reviewer program and approved by the Emergency Preparedness Manager prior to implementation. On an annual basis, written agreements with outside support organizations and government agencies shall be evaluated to determine if such agreements are still valid. (Protected: Ref. FPL Common Letter L-2005-214)

If not, then these agreements shall be renewed and updated ; otherwise, the agreements shall be considered current. Telephone number listings associated with the Station emergency response facilities shall be reviewed quarterly and updated if necessary. Revisions shall be made in accordance with current regulations and guidelines on a continuing basis, as applicable.

Revisions and changes to the plan and procedures shall be forwarded to all document control list recipients. (Protected: Ref. NRC IR 86-18[31])

12.4 Maintenance and Inventory of Emergency Equipment and Supplies Emergency equipment and supplies are maintained as indicated in the Emergency Preparedness Facility Inventory Manual. Emergency portable survey instruments and dosimetry will be calibrated in accordance with applicable health physics programs and procedures. Along with requirements for calibration, the instruments shall be source-checked before each use. There are sufficient reserve instruments and equipment to replace those that are removed from emergency kits for calibration purposes. An inventory of the emergency storage locations shall be made, and discrepancies shall be noted and corrected.

12.5 Emergency Preparedness Manager The Emergency Preparedness Manager is the emergency planning coordinator with overall authority for radiological emergency response planning for Seabrook Station. The Emergency Preparedness Manager has the following responsibilities:

1. Maintain the Seabrook Station Radiological Emergency Plan (SSREP).

2. Maintain the Emergency Response Manual (SSER).

3. Ensure the conduct of drills and exercises.

4. Track identified drill and exercise deficiencies, and associated corrective action.

5. Maintain Emergency Response Organization staffing.

1-12.7 SSREP Rev. 71

6. Maintain Emergency Response Organization pager assignments and publish schedules.

7. Maintain the Emergency Response Organization notification system data base.

8. Maintain the emergency response facilities as described in the Seabrook Station Radiological Emergency Plan and Emergency Response Manual.

9. Obtain and track the availability of facilities and equipment required to maintain the Seabrook Station emergency response in a continuous state of readiness.

10. Ensure implementation of the communications and equipment test program.

12.6 Technical Training Supervisor Ensures the conduct and documentation of emergency preparedness training.

12.7 Operations Support Manager

1. Maintains Operations Department fire response and medical emergency response procedures.

2. Ensures the conduct of fire and medical emergency response drills.

1-12.8 SSREP Rev. 71

13.0

SUMMARY

OF CHANGES Rev. 73: (PCR2217534 August2017)

Section 5 - Corrected footers on EAL charts. Corrected typographical error on Figure 5. 7 (Category A to Category R) (AR 2217534).

Section 12 - Replaced validation, exemption, and deferral are discussed in the ERO Training Program Description with training requirements are discussed in the ERO Training Program Description (AR 2194476).

Appendix D - Updated Letter of Agreement with the State of New Hampshire and Commonwealth of Massachusetts.

Rev. 72: (PCR 2191568 July 2017)

Section 5 - Editorial Change to correct footer in Figures 5.6, 5.7, and 5.8.

Section 9 - Replaced Site Vice President title with Plant General Manager. Corrected typographical errors in Figure 9 .1.

Rev. 71: (PCR 2211605 July 2017)

Section 5 - Revised EAL description to match new NRC approved EAL scheme (AR 2101091 ).

Editorial change to replace reference from NARC to LI-AA-102-1001 with regard to regulatory reporting.

Rev. 70: (AR 02131023 December 2016)

Section 6 - Updated description of new seismic monitor alarms (EC 282184).

Section 8.3 - Removed reference to cancelled STMM and replaced with corporate communications policies.

Figure 8.1 - Reformatted Figure. Added EMT position to more clearly indicate that an EMT is required on shift. Revised note to state that the qualified EMT may be staffed by a member of the Fire Brigade.

Section 10.1.2-Improved description of field monitoring instrumentation sensitivity. Changed terms for TLD to Dosimeter of Legal Record (DLR) and electronic dosimetry to self-reading dosimetry to support NEI Efficiency Bulletin l 6-26c (AR 02168330).

Sections 12.1 - Replaced reference to the site EP Drill and Exercise Manual with applicable fleet and site procedures.

Section 12.1.2.2 - Corrected referenced step number.

Appendix A, Rev. 64 - Returned the Security Personnel position to Table 1. This position was inadvertently deleted due to an editing error during a previous revision (AR 2092861).

1-13.l SSREP Rev. 72

Appendix D, Rev. 61 -Replaced outdated INPO Letter of Agreement with current letter from INPO Website. Updated Remote Monitoring Area LOA with new lease information. Updated Seabrook Fire Department Letter of Agreement.

1-13.2 SSREP Rev. 72

APPENDIXD LETTERS OF AGREEMENT WITH EMERGENCY RESPONSE ORGANIZATIONS SSREP Rev. 62

APPENDIXD LETTERS OF AGREEMENT TABLE OF CONTENTS Date of Agreement

1. Exeter Hospital January 2012

2. Portsmouth Land Acquisition (Remote Monitoring Area) June 2016

3. Wentworth-Douglass Hospital February 2004

4. Seabrook Fire Department June 2016

5. State of New Hampshire and Commonwealth of Massachusetts July 2017

6. Institute of Nuclear Power Operations See NOTE I

7. Portsmouth Police Department May 2013

8. Pease Development Authority (EOF) May 2013

9. Alternate EOF location for beyond Design Basis Events April 2014 (Protected: Ref. NRC IR 85-32[7])

(Protected: Ref. NRC IR 85-32[12])

(Protected : Ref. FPL Common Letter L-2005-214)

NOTE 1: The INPO emergency assistance agreement is initiated by INPO with its member utilities.

The agreement is certified to remain in effect annually by INPO by letter of agreement to its member utilities. The current letter of certification is posted annually by INPO on the INPO website under emergency preparedness. For that reason, the current INPO letter of agreement is not maintained in the SSREP.

SSREP Rev. 62

EMERGENCY RESPONSE PLAN AGREEMENT BY THE STATE OF NEW HAMPSHIRE,

• * . THE COMMONWEALTH OF MASSACHUSETTS AND NextEra ENERGY SEABROOK I. PURPOSE TI1is agreement establishes cooperative anangements for emergency preparedness, notification and response among NextEra Energy Seabrook (NextEra Seabrook) for the Seabrook Station Emergency Response Organization (ERO), tbe State ofNew Hampshire and the Commonwealth of Massachusetts in the event ofa radiological emergency at the Seabrook Station Nuclear Power Plant.

ll. DEFINITIONS A. Emergency Operations Centers (EOCs)- State and local facilities established in NH and MA where emergency response command and control occms.

B. Emergency Operations Facility (EOF) - A NextEra Seabrook offsite emergency response facility where Seabrook Station ERO and offsite emergency response operations are coordinated.

C. EOF Coordinator - The Seabrook Station ERO position that coordinates accident assessment and protective action recommendations with offsite authorities either al the EOF or at other offsite locations such as the State EOCs.

D. Incident Field Office (IFO)-A State ofNew Hampshire facility co-located with the EOF where State ofNew Hampshire response and assistance to EPZ communities can be coordinated.

E. Joint Information Center (TIC) - A facility co-located with the EOF designated for news media briefings conducted jointly by Seabrook Station ERO, State of New Hampshire, Commonwealth of Massachusetts and federal response officials.

F. Nuclear Alert System (NAS) - A dedicated communication system for notification to the State ofNew Hampshire and the Commonwealth of Massachusetts by the Seabrook Station ERO of an emergency and for coordination of public protective action recommendations.

JU. AGREEMENT A. The Seabrook Station ERO shall notify the New Hampshire State Police Dispatch, or other State Warning Point designated by the State of New Hampshire, and the Massachusetts Emergency Management Agency within fifteen (15) minutes after an event has been classified as an Unusual Event, Alert, Sile Area Emergency or General Emergency. This notification shall be made via the NAS. If the NAS is inoperable, notification will be made via a back-up wireless communication network or by commercial telephone.

D-5 SSREP Rev. 57

ill. AGREEMENT (con't)

B. The message content used in the notification identified in A. above is in agreement among the emergency response procedures of the three organizations. After the initial notification, additional information shall be provided to the NH Public Health Emergency Response Initiator and the MA Department of Public Health (MDPH) Radiation Control Program NJ.AT Contact when each calls back to the Seabrook Station ERO. This additional information will include verification of the initial message content, a brief description of events and any known prognosis.

C. NextEra Seabrook shall provide space for representatives from the State ofNew Hampshire and the Commonwealth of Massachusetts at the EOF and the nc. This includes space at the EOF for the State ofNew Hampshire Incident Field Office, Accident Assessment, and for representatives of the Massachusetts Emergency Management Agency and the MDPH Radiation Control Program.

Storage space will also be available for additional radiological equipment for the State of New Hampshire in the IFO.

D. When the EOF is activated by the Seabrook Station ERO, the EOF Coordinator is the point of contact for the offsite representatives at the EOF, or at their respective State EOCs, for plant radiological assessment and fommlation of protective action recommendations. The ERO Technical Liaison is the point of contact at the EOF for plant technical information. NextEra Seabrook will assign liaisons, as available, from Seabrook Station to the Massachusetts and New Hampshire State EOCs to facilitate communication and interpretation of plant technical information.

E. The three organizations agree to exchange information (e.g., radiological releases, meteorological data, offsite dose projections and radiological measurements, and plant technical information) known and available at the time to facilitate a rapid and accurate evaluation of emergency conditions.

F. The State of New Hampshire and the Commonwealth of Massachusetts agree to the methodology used by the Seabrook Station ERO to project offsite radiological consequences. The states may also utilize their own independent dose assessment methods to project and/or measure offsite radiological consequences. The Seabrook Station ERO, including Seabrook Station liaisons at the respective State EOCs, will respond to requests for "what it" projections that are within the capability of the methodology used by the Seabrook Station ERO.

G. The tlu*ee organizations will coordinate offsite field radiological monitoring activities. This coordination is to include the deployment of each organization's offsite radiological monitoring teams (i.e., the states' teams within their respective po1tions of the plume exposure pathway EPZ and the Seabrook Station ERO teams throughout the EPZ), the review and exchange of all monitoring results, and the radionuclide analysis ofpmticulate and radio-iodine samples at the EOF.

D-5a SSREP Rev. 57

ill. AGREEMENT (con't)

H. NextEra Seabrook agrees to support the State ofNew Hampshire and the Commonwealth of Massachusetts with ingestion pathway sampling and analysis. Th.is support will include coordination of ingestion pathway sampling plans, coordination of ingestion pathway sample analysis, and independent ingestion pathway dose calculations based on sample analysis results.

I. NextEra Seabrook agrees to make available to the states the services of any radio-analysis laboratories maintained by, contracted by, or otherwise available to NextEra Seabrook for use by the Seabrook Station ERO in the event of a radiological emergency. The State of New Hampshire and the Commonwealth of Massachusetts agree to make available to the Seabrook Station ERO radio-analysis laboratory services maintained by, contracted by or otherwise available to the states for use in the event of a radiological emergency. Sample processing priorities will be established through joint agreement by the Seabrook Station ERO dose assessment staff, NH Public Health personnel, and MDPH Radiation Control Program personnel.

J. The Commonwealth of Massachusetts and the State ofNew Hampshire agree to coordinate the evaluation and implementation ofprecautionaiy actions for special populations and persons with special needs within the plume exposure pathway EPZ.

K. The Commonwealth of Massachusetts and the State of New Hampshire agree to coordinate the evaluation and implementation of protective actions for the general public and the notification of the public by activation of the public alert and notification system, includ.ing activation of the Emergency Alert System.

L. The three organizations agree to coordinate news statements and media briefings at the JIC. The states will staff rumor control telephones at their respective EOCs, or other designated site(s), and will relay rumor trends to the JIC to be addressed via the news media. Information shall be made available to the public in a timely, coordinated manner through the JIC and/or other agreed upon means.

M. The Commonwealth of Massachusetts and the State of New Hampshire have reviewed and agree to the procedure utilized by the Seabrook Station ERO to classify emergencies, including the Seabrook Station Emergency Action Levels (EALs). NextEra Seabrook will review substantive changes made to EALs with appropriate New Hampshire and Massachusetts state emergency management officials. NextEra Seabrook staff will document this review.

N. The State of New Hampshire agrees to notify the State of Maine of all emergency classification levels declared at Seabrook Station.

0. The State ofNew Hampshire agrees to notify and to coordinate response actions with the Un.ited States Coast Guard for the off-shore waterway portions of the plume exposure pathway EPZ.

P. The State ofNew Hampshire agrees to notify the Federnl Aviation Admin.istration for air space restrictions over the plwne exposure pathway EPZ.

Q. The State of New Hampshire and Commonwealth of Massachusetts agree to notify passenger and freight rail services originating within their respective jurisdiction of rail travel restrictions within the plume exposure pathway EPZ. The Stale ofNew Hampshire agrees to notify the State of Maine regarding passenger and freight rail services restrictions within the plume exposure pathway EPZ for rail services originating within the State of Maine.

D-5b SSREP Rev. 57

lil. AGRimMENT (con't)

R. The Commonwealth of Massachusetts and the State of New Hampshire agree to arrange for clearance of Seabrook Stat.ion ERO and emergency response support personnel through EPZ access control points.

S,- 111e three organizations shall exchange and agree to radiological emergency response plan changes that pe1tain to notification of an emergency and to coordination of emergency response actions prior to implementing such changes.

T. 111is agreement may be amended at any time by written agreement among the parties.

U. This agreement shall be effective as of the date of the last s.ignature shown below.

STATE OF NEW HAMPSHIRE COMMON~OF MASSACHUSEITS By A ti!Jmi?<J By I lt~ ;.,,,...,:-r::_

/foh~Barthelmes Kurt Schwartz Director Commissioner Massachusetts Emergency NH Department of Safety 011a__ M~z:J,~;

, DATE By O./})Jq+--

ffrey A. Meyers Commissioner NH Department of Health &

Human Services

--~lb_ DATE NEXTERA ENERGY SEABROOK B~~ liCMCcartn Rcgiona Vice President

\

D-5c SSREP Rev. 57

SUMMARY

OF CHANGES Rev. 62: (PCR 2217534 August 2017)

Updated Appendix D with the revised letter of agreement between the site and the State of New Hampshire and Commonwealth of Massachusetts.

Rev. 61:

In Appendix D updated agreements for Portsmouth Land Acquisition (Remote Monitoring Area),

Seabrook Fire Department, and Institute of Nuclear Power Operations.

Rev. 60:

In Appendix D added agreement for alternative EOF for beyond design basis events.

Rev. 59:

In Appendix D removed PSNH Newington Station and Newington Police Department. Added Portsmouth Police Department and Pease Development Authority. Changed Jask Realty to Portsmouth Land Acquisition. (AR# 1721945)

Rev. 58:

Added new lease agreement for relocated remote monitoring area and updated letters of agreement for Exeter Hospital and Seabrook Fire Department.

Rev. 57:

Inserted most recent Institute of Nuclear Power Operations emergency assistance agreement.

Rev. 56:

Updated letter of agreement with the Commonwealth of Massachusetts and the State of New Hampshire.

Updated lease agreements with Public Service New Hampshire for the Emergency Operations Facility at Newington Station and for the remote monitoring/decontamination facility at Schiller Station.

Rev. 55:

Updated the letter of agreement with the Seabrook Fire Department with agreement dated August 2007.

Rev. 54:

This appendix was unaffected by this revision to the manual.

D-10 SSREP Rev. 62

SEABROOK STATION ADMINISTRATIVE PROCEDURE Classification of Emergencies ER 1.1 Rev. 58 Procedure Owner:

D. Currier

ER 1.1 Page 2 Rev. 58 Contents and Revision Status Contents Page No.

Cover 1 Contents and Revision Status 2 1.0 OBJECTIVES 3 1.1 Discussion 3 2.0 RESPONSIBILITIES 6 2.1 Unit Supervisor 6 2.2 Shift Manager 6 2.3 Site Emergency Director 6 3.0 PRECAUTIONS 6 4.0 PREREQUISITES 7 5.0 ACTIONS 7 5.1 Emergency Classification 7

6.0 REFERENCES

9 7.0 ATTACHMENTS Figure 1 Initiating Conditions and Emergency Action Levels 9 Figure 2 Definitions 133 Figure 3 Acronyms and Abbreviations 136 Figure 4 Additional Basis Information 138 Figure 5 Summary of Changes 148 Rev. No.

ER 1.1A EMERGENCY INITIATING CONDITION MATRIX 47 Modes 1, 2, 3 and 4 ER 1.1B EMERGENCY INITIATING CONDITION MATRIX 47 Modes 5, 6 and Defueled ER 1.1C FISSION PRODUCT BARRIER DEGRADATION MATRIX 47 Modes 1, 2, 3 and 4

ER 1.1 Page 3 Rev. 58 1.0 OBJECTIVES This procedure specifies the Initiating Conditions and Emergency Action Levels (EALs) used to classify emergencies in accordance with the Seabrook Station Radiological Emergency Plan.

1.1 Discussion When making classifications based upon lights, alarms and other indications, caution must be exercised in the review of these indications to ensure the validity of the information.

When two or more emergency initiating conditions exist and different emergency classifications result, the higher emergency classification will be used.

Critical Safety Function Status Tree Critical safety function evaluation for emergency classification purposes should be performed in accordance with the guidance contained in the Operations Management Manual (OPMM). Critical safety functions for emergency classification considerations apply when they become applicable in the EOP network. In Modes 1, 2 and 3, this is defined as after leaving E-0 or when directed by E-0 to monitor them. In Mode 4, an emergency classification based on a non-green Critical Safety Function Status Tree (CSFST) would be required if the CSFST is applicable and operators make an entry into the associated Functional Restoration Procedure (FRP).

Proper use of the CSFSTs as EALs requires that CSF challenges be identified consistent with the procedure transition criteria specified in the EOP network. A non-green CSFST should not be used as the basis for emergency classification unless the EOP criteria directing the transition into the associated FRP are met. This includes situations where the operation of plant equipment, in accordance with an EOP network procedure, is driving a non-green CSFST terminus. A CSF must be "truly" challenged, in a sense recognized by the EOP network procedures, for the associated CSFST to be the basis for a valid emergency declaration.

An emergency declaration in the Control Room occurs when the Short Term Emergency Director (STED) announces the selected Emergency Class to the Control Room staff in accordance with the applicable STED checklist in Procedure ER 1.2, Emergency Plan Activation. An emergency declaration in the Technical Support Center (TSC) occurs when the Site Emergency Director (SED) announces the selected Emergency Class to the TSC staff. If the SED is in the Control Room during event reclassification, the emergency declaration occurs when the SED announces the selected Emergency Class to the Control Room staff.

Nuclear power reactor licensees shall 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 shall promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. Licensees shall not construe these criteria as a grace period to attempt to restore plant conditions to avoid declaring an emergency due to an emergency action level that has been exceeded. Licensees shall not construe these criteria as preventing implementation of response actions deemed by the licensee to be necessary to protect public health and safety

ER 1.1 Page 4 Rev. 58 provided that any delay in declaration does not deny the state and local authorities the opportunity to implement measures necessary to protect public health and safety.

Certain transient conditions mask indications to plant operators. For example, a significant steam leak contributing to a plant cool down may be of sufficient magnitude to impact the ability to diagnose and quantify primary leakage. In this case operators are expected to exhaust all alternate means of indication to determine if an EAL has been exceeded. IF no other conclusive indication is present then "availability of indications to plant operators that an emergency action level has been exceeded" is not satisfied. (AR 1986114).

For EAL thresholds that specify a duration of the off-normal condition (e.g., 15 minutes), the emergency declaration "clock" runs concurrently with the specified threshold duration "clock".

Once the off-normal condition has existed for the duration specified in the EAL, no further classification assessment time is allowed. The EAL has been exceeded and a prompt emergency declaration is required. Example: Initiating Condition HU4, FIRE not extinguished within 15 minutes following direct report, multiple fire alarms, or field verification of a fire alarm. On receipt of a fire indication, the Fire Brigade is dispatched to the scene to begin fire fighting efforts. If the fire is still burning after 15 minutes has elapsed, the EAL is exceeded and the emergency declaration must be made promptly.

If emergency conditions are initially classified as an Alert or higher, and then subsequently reclassified to an Unusual Event, all ERO members should continue to report to their facilities.

Although activation of the Technical Support Center, Operational Support Center, and Emergency Operations Facility are not required, the ERO staff will be available to assist with event recovery efforts, interface with State emergency response personnel, and respond to information requests from the media, elected officials and industry organizations.

Standard emergency class definitions are as follows:

• Unusual Event - Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

• Alert - Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

• 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.

ER 1.1 Page 5 Rev. 58

• 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.

2.0 RESPONSIBILITIES 2.1 Unit Supervisor Responsible for assuming the role of Short Term Emergency Director (STED) until the Shift Manager has reported to the Control Room.

2.2 Shift Manager Responsible for classifying observed station conditions in accordance with the emergency classification system specified in this procedure and reclassifying the emergency as necessary until relieved by the Site Emergency Director.

2.3 Site Emergency Director Responsible for analyzing changing station conditions and reclassifying the emergency classification in accordance with this procedure.

3.0 PRECAUTIONS

1. Final emergency classifications are contingent upon the evaluation and discretion of the Shift Manager or the Site Emergency Director. The Shift Manager or Site Emergency Director may make an emergency classification based on clear indications that the event trajectory meets the intent of the initiating condition, although the associated emergency action levels have not yet been met or exceeded.

2. Critical safety function status tree (CSFST) color displays must be sustained indications of continuous conditions. Conditions indicated by CSFST displays must be evaluated and verified using hardwired information before they are used as bases for emergency classifications or for protective action recommendations.

ER 1.1 Page 6 Rev. 58

3. Offsite dose projections are required in the event that any of the following conditions occur:

a. HI alarm on Wide Range Gas Monitor (WRGM) effluent rate monitor (RM-6528-4), or

b. HI alarm on a Main Steam Line Monitor with an OPEN atmospheric steam dump valve (ASDV) or safety relief valve (SRV) on the affected line, or

c. HI alarm on a Main Steam Line Monitor with the steam driven EFW pump running and fed from the affected line.

At the discretion of the Shift Manager, offsite dose projections may be performed after the initial declaration is made based on other plant or radiological conditions.

4. An emergency declaration should be made as soon as possible after indications are available that an EAL has been exceeded, not to exceed 15 minutes, unless delay is required by response actions necessary to protect public health and safety.

4.0 PREREQUISITES An emergency initiating condition has occurred.

5.0 ACTIONS Shift Manager/Site Emergency Director 5.1 Emergency Classification

1. Depending upon the plant mode in effect when the potential emergency initiating condition occurs, review the following forms.

Station Mode Forms to Review 1, 2, 3 or 4 ER 1.1A, Emergency Initiating Condition Matrix Modes 1, 2, 3, and 4 ER 1.1C, Fission Product Barrier Degradation Matrix Modes 1, 2, 3, and 4 5, 6 and Defueled ER 1.1B, Emergency Initiating Condition Matrix Modes 5, 6, and Defueled

2. If an emergency classification is being considered under a radiological effluent EAL which requires a dose projection (RS1 or RG1), implement offsite dose assessment using procedure ER 5.3, Operation of the Raddose-V.

In the event of a radiological release via the turbine-driven EFW pump exhaust, dispatch a monitoring team to the downwind site boundary location to obtain a site boundary dose rate and use the Unmonitored Release Path of Raddose.

ER 1.1 Page 7 Rev. 58

3. Circle the potential emergency initiating condition(s) on each Form. This assessment must be performed promptly to support the goal of making an emergency declaration within 15 minutes of initial EAL indications becoming available in the Control Room.

4. For Category R, E, H, M and C events, refer to the initiating condition EAL(s) in Figure 1 and verify that either the EAL(s) is met or the intent is met. All Category F EALs are presented on Form ER 1.1C (i.e., not in Figure 1).

5. Identify the most severe (highest) emergency classification for which the EAL(s) is met or the intent of the initiating condition is met.

6. If the emergency-related indications completely clear before a declaration of an emergency classification level has been made, then no emergency classification is required.

• The Shift Manager shall notify the Emergency News Manager within one hour of the termination of the emergency-related indications that emergency-related indications briefly existed, but cleared prior to the declaration of an emergency classification.

• Direct the Emergency News Manager to initiate state notifications per good neighbor notification procedures.

• The event shall be reported to the NRC in accordance with 10 CFR 50.72 and 50.73 per LI-AA-102-1001, Regulatory Reporting, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the discovery of the undeclared event.

7. If emergency-related indications are received and later cleared, and after the fact it is determined that an emergency classification was warranted but not made, then no emergency classification is required.

• The Shift Manager shall notify the Emergency News Manager within one hour of discovery that an emergency classification was warranted but not declared and that emergency-related indications no longer exist.

• Direct the Emergency News Manager to initiate state notifications per good neighbor notification procedures.

• The event shall be reported to the NRC in accordance with 10 CFR 50.72 and 50.73 per the LI-AA-102-1001, Regulatory Reporting, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the discovery of the undeclared event.

8. If emergency-related indications are received and reduce in severity, such that the emergency classification went from an earlier higher level to a current lower level, the current lower level emergency should be declared. State and NRC notifications shall be made in accordance with Procedure ER 1.2.

ER 1.1 Page 8 Rev. 58 NOTE Steps 9 and 10 are not applicable after the Technical Support Center is activated at an Alert or higher classification.

9. If an emergency declaration is warranted, immediately implement Station Emergency Response Procedure ER 1.2, Emergency Plan Activation.

10. If an emergency declaration is not warranted, document the applicability review and basis for the decision not to implement the emergency plan in the Unit journal.

6.0 REFERENCES

1. Seabrook Station Radiological Emergency Plan

2. ER 1.2, Emergency Plan Activation

3. ER 5.3, Operation of the Raddose-V

4. E-3, Steam Generator Tube Rupture

Figure 1 ER 1.1 Page 9 Initiating Conditions and Emergency Action Levels Rev. 58 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY RG1 Release of RS1 Release of RA1 Release of RU1 Release of gaseous radioactivity gaseous radioactivity gaseous or liquid gaseous or liquid resulting in offsite resulting in offsite radioactivity resulting radioactivity greater dose greater than 1,000 dose greater than 100 in offsite dose greater than 2 times the mrem TEDE or 5,000 mrem TEDE or 500 than 10 mrem TEDE ODCM limits for 60 mrem thyroid CDE. mrem thyroid CDE. or 50 mrem thyroid minutes or longer.

Op. Modes: All Op. Modes: All CDE. Op. Modes: All Op. Modes: All RG2 Spent fuel pool RS2 Spent fuel pool RA2 Significant RU2 UNPLANNED level cannot be level at 1.5 ft. (Level lowering of water level loss of water level restored to at least 1.5 3). above, or damage to, above irradiated fuel.

ft. (Level 3) for 60 Op. Modes: All irradiated fuel. Op. Modes: All minutes or longer. Op. Modes: All Op. Modes: All RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.

Op. Modes: All

Figure 1 ER 1.1 Page 10 Initiating Conditions and Emergency Action Levels Rev. 58 RG1 RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)

Notes:

The STED/SED should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

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

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

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

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:

Monitor Reading RM-6528-4 (WRGM rate) 2.85E+8 uCi/sec Time After Shutdown Reading 1 hr > 1 hr to 2 hrs RM-6481-1* (MSL A) 1310 mR/hr 1060 mR/hr RM-6482-1* (MSL B) 1310 mR/hr 1060 mR/hr RM-6482-2* (MSL C) 1310 mR/hr 1060 mR/hr RM-6481-2* (MSL D) 1310 mR/hr 1060 mR/hr

• With release path to the environment from affected steam line, open ASDV or SRV, line is faulted, or open steam supply to 1-FW-P-37A.

• • OR **

(2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.

• • OR **

(3) Field survey results indicate EITHER of the following at or beyond the site boundary:

Closed window dose rates greater than 1000 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 5000 mrem for one hour of inhalation.

Figure 1 ER 1.1 Page 11 Initiating Conditions and Emergency Action Levels Rev. 58 RG1 BASIS INFORMATION RG1 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.

Figure 1 ER 1.1 Page 12 Initiating Conditions and Emergency Action Levels Rev. 58 RG2 RG2 ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least 1.5 ft. (Level 3) for 60 minutes or longer.

Operating Mode Applicability: All Emergency Action Levels:

Note:

The STED/SED should declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.

(1) Spent fuel pool level cannot be restored to at least 1.5 ft. above the fuel racks for 60 minutes or longer as indicated by SF-LI-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point A4220).

Figure 1 ER 1.1 Page 13 Initiating Conditions and Emergency Action Levels Rev. 58 RG2 BASIS INFORMATION RG2 Basis:

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

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SFP level 1.5 ft. above the top of the fuel racks (Level 3).

The Spent Fuel Pool Instrumentation System (SFPIS) consists of two new independent level instrument channels (SF-L-2616 and SF-L-2617) in the Spent Fuel system. The SFPIS channels will be used to monitor spent fuel pool level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.

Each channel is capable of measuring SFP level over a span from just above the top of the spent fuel racks to the normal SFP operating water level. The SFPIS will be monitored in accordance with Beyond Design Basis guidelines contained in FSGs for Extended Loss of AC Power and Alternate SFP Makeup and Cooling.

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

Figure 1 ER 1.1 Page 14 Initiating Conditions and Emergency Action Levels Rev. 58 RS1 RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)

Notes:

The STED/SED should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

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

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

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

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:

Monitor Reading RM-6528-4 (WRGM rate) 2.85E+7 uCi/sec Time After Shutdown Reading 1 hr > 1 hr to 2 hrs RM-6481-1* (MSL A) 130 mR/hr 100 mR/hr RM-6482-1* (MSL B) 130 mR/hr 100 mR/hr RM-6482-2* (MSL C) 130 mR/hr 100 mR/hr RM-6481-2* (MSL D) 130 mR/hr 100 mR/hr

• With release path to the environment from affected steam line, open ASDV or SRV, line is faulted, or open steam supply to 1-FW-P-37A.

• • OR **

(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.

• • OR **

(3) Field survey results indicate EITHER of the following at or beyond the site boundary:

Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Figure 1 ER 1.1 Page 15 Initiating Conditions and Emergency Action Levels Rev. 58 RS1 BASIS INFORMATION RS1 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.

Figure 1 ER 1.1 Page 16 Initiating Conditions and Emergency Action Levels Rev. 58 RS2 RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at 1.5 ft.(Level 3)

Operating Mode Applicability: All Emergency Action Levels:

(1) Lowering of spent fuel pool level to 1.5 ft. above the fuel racks as indicated by SF-LI-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point A4220).

Figure 1 ER 1.1 Page 17 Initiating Conditions and Emergency Action Levels Rev. 58 RS2 BASIS INFORMATION RS2 Basis:

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

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SFP level 1.5 ft. above the top of the fuel racks (Level 3).

The Spent Fuel Pool Instrumentation System (SFPIS) consists of two new independent level instrument channels (SF-L-2616 and SF-L-2617) in the Spent Fuel system. The SFPIS channels will be used to monitor spent fuel pool level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.

Each channel is capable of measuring SFP level over a span from just above the top of the spent fuel racks to the normal SFP operating water level. The SFPIS will be monitored in accordance with Beyond Design Basis guidelines contained in FSGs for Extended Loss of AC Power and Alternate SFP Makeup and Cooling.

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.

Figure 1 ER 1.1 Page 18 Initiating Conditions and Emergency Action Levels Rev. 58 RA1 RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)

Notes:

The STED/SED should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

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

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

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

(1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:

Monitor Reading RM-6528-4 (WRGM rate) 2.85E+6 uCi/sec RM-6481-1* (MSL A) 10 mR/hr RM-6482-1* (MSL B) 10 mR/hr RM-6482-2* (MSL C) 10 mR/hr RM-6481-2* (MSL D) 10 mR/hr

• With release path to the environment from affected steam line, open ASDV or SRV, line is faulted, or open steam supply to 1-FW-P-37A.

• • OR **

(2) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary.

• • OR **

(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary for one hour of exposure.

• • OR **

(4) Field survey results indicate EITHER of the following at or beyond the site boundary:

Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

Figure 1 ER 1.1 Page 19 Initiating Conditions and Emergency Action Levels Rev. 58 RA1 BASIS INFORMATION RA1 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.

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 RS1.

Figure 1 ER 1.1 Page 20 Initiating Conditions and Emergency Action Levels Rev. 58 RA2 RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)

(1) Uncovery of irradiated fuel in the REFUELING PATHWAY.

• • OR **

(2) Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by high-alarm, or reading in excess of the current high-alarm setpoint on ANY of the following radiation monitors:

RM-6518-1 FSB High Range RM-6562-1 FSB Vent RM-6535A-1 Manipulator Crane RM-6535B-1 Manipulator Crane

• • OR **

(3) Lowering of spent fuel pool level to 12 ft. 3 in. above the fuel racks on SF-LI-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point A4220).

Figure 1 ER 1.1 Page 21 Initiating Conditions and Emergency Action Levels Rev. 58 RA2 BASIS INFORMATION RA2 Basis:

REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal.

This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. 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 EU1.

Escalation of the emergency would be based on either Recognition Category R or C ICs. EAL #1 This EAL escalates from RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation, as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used.

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. 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.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 22 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION RA2 RA2 CONTINUED 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.

Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SFP level 1.5 ft. above the top of the fuel racks (Level 3).

The Spent Fuel Pool Instrumentation System (SFPIS) consists of two new independent level instrument channels (SF-L-2616 and SF-L-2617) in the Spent Fuel system. The SFPIS channels will be used to monitor spent fuel pool level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.

Each channel is capable of measuring SFP level over a span from just above the top of the spent fuel racks to the normal SFP operating water level. The SFPIS will be monitored in accordance with Beyond Design Basis guidelines contained in FSGs for Extended Loss of AC Power and Alternate SFP Makeup and Cooling.

Escalation of the emergency classification level would be via ICs RS1 or RS2.

Figure 1 ER 1.1 Page 23 Initiating Conditions and Emergency Action Levels Rev. 58 RA3 RA3 ECL: Alert Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2)

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

(1) Dose rate greater than 15 mR/hr in ANY of the following areas:

Control Room RM6550 Central Alarm Station (CAS) by survey Secondary Alarm Station (SAS) by survey

• • OR **

(2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any of the following plant rooms or areas:

Table H1 Area Mode Primary Aux Building 25 ft elevation 1, 2, 3, 4 7 ft elevation

- 26 ft elevation Turbine Building 21 ft elevation 1, 2, 3 50 ft elevation Essential Switchgear Rooms 1, 2, 3, 4 Waste Process Building 25 ft elevation 1, 2, 3

-3 ft elevation Containment 3, 4 RHR/CBS Equipment Vaults 3, 4

Figure 1 ER 1.1 Page 24 Initiating Conditions and Emergency Action Levels Rev. 58 RA3 BASIS INFORMATION RA3 Basis:

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

IMPEDE: Entry into an area requires extraordinary measures to facilitate entry of personnel into the affected room/area by installing temporary shielding, requiring use of non-routine protective equipment, or requesting an extension in dose limits beyond normal administrative limits.

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 STED/SED 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.

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.

• The action for which room/area entry is required is of an administrative or record keeping nature.

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

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

Figure 1 ER 1.1 Page 25 Initiating Conditions and Emergency Action Levels Rev. 58 RU1 RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2)

Notes:

The STED/SED should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.

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

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

1) a Valid reading on ANY of the following effluent monitors greater than 2 times the value of the current high-alarm setpoint for 60 minutes or longer:

RM-6509-1 (WTT Disch)

RM-6521-1 (TB Sump)

RM-6519-1 (SG Blowdown)

RM-6473-1 (WT LIQ EFF)

RM-6528-4 (WRGM rate)

• • AND **

b. The discharge flow to the environment is not isolated within 60 minutes.

• • OR **

2) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.

Figure 1 ER 1.1 Page 26 Initiating Conditions and Emergency Action Levels Rev. 58 RU1 BASIS INFORMATION RU1 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. 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.

EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. EAL #1 addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed two times the ODCM limit and releases are not terminated within 60 minutes. This alarm setpoint may be associated with a planned batch release, or a continuous release path. In either case, the setpoint is established by the ODCM. Indexing the EAL threshold to the ODCM setpoints in this manner insures that the EAL threshold will never be less than the setpoint established by a specific discharge permit.

The discharge flowpaths associated with RM-6509-1, 6521-1, 6519-1, and 6473-1 have automatic and manual flow isolation capability. The EAL wording addresses a situation where a residual source term exists in a discharge flowpath AFTER the flowpath has been isolated, and the associated radiation monitor remains at values above 2 times the value of the current high- alarm setpoint. EAL 1.b ensures that the Initiating Condition (IC) intent of "to the environment" is met. The 60-minute assessment clock starts at the same time for both EAL 1.a and 1.b (i.e., clocks run concurrently). There must be a release to the environment (i.e., the flowpath cannot be isolated) during the same period that a monitor value is greater than 2 times the value of the current high-alarm setpoint.

EAL #2 - This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways.

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

Figure 1 ER 1.1 Page 27 Initiating Conditions and Emergency Action Levels Rev. 58 RU2 RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.

Operating Mode Applicability: All Emergency Action Levels:

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

1-SF-LI-2607 (Spent Fuel Pool Level) 1-SF-LI-2629 or 1-SF-LIT-2629-1 (Reactor Refuel Cavity Level)

• • AND **

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors:

RM-6535-A-1, Containment Manipulator Crane RM-6535-B-1, Containment Manipulator Crane RM-6549-1, FSB Spent Fuel Range Low RM-6518-1, FSB Spent Fuel Range Hi

Figure 1 ER 1.1 Page 28 Initiating Conditions and Emergency Action Levels Rev. 58 RU2 BASIS INFORMATION RU2 Basis:

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

REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. 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 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.

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 RA2.

Figure 1 ER 1.1 Page 29 Initiating Conditions and Emergency Action Levels Rev. 58 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Recognition Category H Initiating Condition Matrix GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY HS1 HOSTILE ACTION HA1 HOSTILE ACTION HU1 Confirmed within the PROTECTED within the OWNER SECURITY CONDITION or AREA. CONTROLLED AREA or threat.

Op. Modes: All airborne attack threat within Op. Modes: All 30 minutes.

Op. Modes: All HU2 Seismic event greater than OBE levels.

Op. Modes: All HU3 Hazardous event.

Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.

Op. Modes: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, shutdown or cooldown.

Op. Modes: All HS6 Inability to control a HA6 Control Room key safety function from evacuation resulting in outside the Control Room. transfer of plant control to Op. Modes: All alternate locations.

Op. Modes: All HG7 Other conditions HS7 Other conditions HA7 Other conditions HU7 Other conditions exist which in the judgment exist which in the judgment exist which in the judgment exist which in the judgment of the STED/SED warrant of the STED/SED warrant of the STED/SED warrant of the STED/SED warrant declaration of a General declaration of a Site Area declaration of an Alert. declaration of an Unusual Emergency. Emergency. Op. Modes: All Event.

Op. Modes: All Op. Modes: All Op. Modes: All

Figure 1 ER 1.1 Page 30 Initiating Conditions and Emergency Action Levels Rev. 58 HG7 HG7 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the STED/SED warrant declaration of a General Emergency.

Operating Mode Applicability: All Emergency Action Levels:

(1) Other conditions exist which in the judgment of the STED/SED 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.

HG7 BASIS INFORMATION HG7 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 STED/SED to fall under the emergency classification level description for a General Emergency.

Figure 1 ER 1.1 Page 31 Initiating Conditions and Emergency Action Levels Rev. 58 HS1 HS1 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.

Operating Mode Applicability: All Emergency Action Levels:

Note: This Initiating Condition and EAL do not apply to an attack solely on the Dry Fuel Storage Protected Area. An attack on the Dry Fuel Storage Facility Protected Area should be considered an attack within the Owner Controlled Area and classified as an Alert per Initiating Condition HA1.

(1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by security shift supervision

Figure 1 ER 1.1 Page 32 Initiating Conditions and Emergency Action Levels Rev. 58 HS1 BASIS INFORMATION HS1 Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP.

PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan.

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures. The Site Area Emergency declaration will mobilize ORO resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

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 Plan.

Escalation of the emergency classification level would be via IC HG7.

Figure 1 ER 1.1 Page 33 Initiating Conditions and Emergency Action Levels Rev. 58 HS6 HS6 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.

Operating Mode Applicability: All Emergency Action Levels:

Note: The STED/SED should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) a. An event has resulted in plant control being transferred from the Control Room to the Remote Safe Shutdown components.

• • AND **

b. Control of ANY of the following key safety functions is not reestablished within 15 minutes.

Reactivity control Core cooling RCS heat removal HS6 BASIS INFORMATION HS6 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 STED/SED judgment. The STED/SED 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).

Escalation of the emergency classification level would be via IC FG1 or CG1.

Figure 1 ER 1.1 Page 34 Initiating Conditions and Emergency Action Levels Rev. 58 HS7 HS7 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the STED/SED warrant declaration of a Site Area Emergency.

Operating Mode Applicability: All Emergency Action Levels:

(1) Other conditions exist which in the judgment of the STED/SED 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 HS7 BASIS INFORMATION HS7 Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the STED/SED to fall under the emergency classification level description for a Site Area Emergency.

Figure 1 ER 1.1 Page 35 Initiating Conditions and Emergency Action Levels Rev. 58 HA1 HA1 ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2)

(1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA or the Dry Fuel Storage Facility as reported by security shift supervision.

• • OR **

(2) A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.

Figure 1 ER 1.1 Page 36 Initiating Conditions and Emergency Action Levels Rev. 58 HA1 BASIS INFORMATION HA1 Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP.

OWNER CONTROLLED AREA: The site property owned by, or otherwise under the control of, the licensee.

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. The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

EAL #1 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.

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 site procedures.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 37 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION HA1 HA1 CONTINUED The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Escalation of the emergency classification level would be via IC HS1.

Figure 1 ER 1.1 Page 38 Initiating Conditions and Emergency Action Levels Rev. 58 HA5 HA5 ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, shutdown or cooldown.

Operating Mode Applicability: All Emergency Action Levels:

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

(1) a. Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H1 rooms or areas.

• • AND **

b. Entry into the room or area is prohibited or IMPEDED.

Table H1 Area Mode Primary Aux Building 25 ft elevation 1, 2, 3, 4 7 ft elevation

- 26 ft elevation Turbine Building 21 ft elevation 1, 2, 3 50 ft elevation Essential Switchgear Rooms 1, 2, 3, 4 Waste Process Building 25 ft elevation 1, 2, 3

-3 ft elevation Containment 3, 4 RHR/CBS Equipment Vaults 3, 4

Figure 1 ER 1.1 Page 39 Initiating Conditions and Emergency Action Levels Rev. 58 HA5 BASIS INFORMATION HA5 Basis:

IMPEDE: Entry into an area requires extraordinary measures to facilitate entry of personnel into the affected room/area by installing temporary shielding, requiring use of non-routine protective equipment, or requesting an extension in dose limits beyond normal administrative limits.

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 STED/SEDs 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.

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.

• The action for which room/area entry is required is of an administrative or record keeping nature.

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

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 40 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION HA5 HA5 CONTINUED An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that 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.

References:

OS1000.03, Plant Shutdown From Minimum Load to Hot Standby OS1000.04, Plant Cooldown From Hot Standby to Cold Shutdown

Figure 1 ER 1.1 Page 41 Initiating Conditions and Emergency Action Levels Rev. 58 HA6 HA6 ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.

Operating Mode Applicability: All Emergency Action Levels:

(1) Entry into Procedure OS1200.02 for control room evacuation resulted in plant control being transferred from the Control Room to Remote Safe Shutdown components.

Figure 1 ER 1.1 Page 42 Initiating Conditions and Emergency Action Levels Rev. 58 HA6 BASIS INFORMATION HA6 Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Escalation of the emergency classification level would be via IC HS6.

Figure 1 ER 1.1 Page 43 Initiating Conditions and Emergency Action Levels Rev. 58 HA7 HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the STED/SED warrant declaration of an Alert.

Operating Mode Applicability: All Emergency Action Levels:

(1) Other conditions exist which, in the judgment of the STED/SED, 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.

Figure 1 ER 1.1 Page 44 Initiating Conditions and Emergency Action Levels Rev. 58 HA7 BASIS INFORMATION HA7 Basis:

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the STED/SED to fall under the emergency classification level description for an Alert.

Figure 1 ER 1.1 Page 45 Initiating Conditions and Emergency Action Levels Rev. 58 HU1 HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)

(1) A Code Yellow is reported by the Security Shift Supervisor.

• • OR **

(2) Notification of a credible security threat directed at Seabrook Station.

• • OR **

(3) A validated notification from the NRC providing information of an aircraft threat.

Figure 1 ER 1.1 Page 46 Initiating Conditions and Emergency Action Levels Rev. 58 HU1 BASIS INFORMATION HU1 Basis:

Code Yellow - 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.

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP.

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG7.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and OROs.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

EAL #1 references Security Shift Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.390 information.

EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with site procedures.

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 site procedures.

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 Plan.

Escalation of the emergency classification level would be via IC HA1.

References:

OS1290.03, Response to a Security Event. OS1290.04, Response to an Airborne Security Event

Figure 1 ER 1.1 Page 47 Initiating Conditions and Emergency Action Levels Rev. 58 HU2 HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels.

Operating Mode Applicability: All Emergency Action Levels:

(1) Seismic event greater than Operating Basis Earthquake (OBE) as indicated by:

a. The red "EVENT" light is lit on seismic monitoring control panel 1-SM-CP-58.

• • AND **

b. The yellow "OBE" light is lit on seismic monitoring control panel 1-SM-CP-58.

• • OR **

(2) a. Seismic monitoring system out of service

• • AND **

b. Control Room personnel feel an actual or potential seismic event

• • AND **

c. The occurrence of a seismic event is confirmed in a manner deemed appropriate by the STED/SED

Figure 1 ER 1.1 Page 48 Initiating Conditions and Emergency Action Levels Rev. 58 HU2 BASIS INFORMATION HU2 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. Given the time necessary to perform walk- downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The STED/SED may seek external verification if deemed appropriate; however, the verification action must not preclude a timely emergency declaration.

Reference:

EC 282184, Seismic Monitoring System Upgrade Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or MA9.

Figure 1 ER 1.1 Page 49 Initiating Conditions and Emergency Action Levels Rev. 58 HU3 HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)

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

(1) A tornado strike within the PROTECTED AREA.

• • OR **

(2) Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.

• • OR **

(3) Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials.

• • OR **

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

Figure 1 ER 1.1 Page 50 Initiating Conditions and Emergency Action Levels Rev. 58 HU3 BASIS INFORMATION HU3 Basis:

PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan.

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. Systems classified as safety- related.

IMPEDE: Entry into an area requires extraordinary measures to facilitate entry of personnel into the affected room/area by installing temporary shielding, requiring use of non-routine protective equipment, or requesting an extension in dose limits beyond normal administrative limits.

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.

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. 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.

Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, M or C.

Figure 1 ER 1.1 Page 51 Initiating Conditions and Emergency Action Levels Rev. 58 HU4 HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant.

Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)

Notes:

• The STED/SED should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

• A containment fire alarm is considered valid upon receipt of multiple zones (more than

1) actuated on CP-376 panel.

(1) a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

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

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

• • AND **

b. The FIRE is located within ANY Table H2 plant rooms or areas:

Table H2 Condensate Storage Tank Enclosure Fuel Storage Building Containment Primary Auxiliary Building Control Building Service Water Pump House Cooling Tower Steam and Feedwater Pipe Chases Diesel Generator Building North Tank Farm Emergency Feedwater Pump House Startup Feedwater Pump Area RHR/CBS Equipment Vaults

• • OR **

(2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).

• • AND **

b. The FIRE is located within ANY of the Table H2 plant rooms or areas except Containment in Modes 1 and 2 (see note above):

• • AND **

c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.

• • OR **

(3) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility not extinguished within 60-minutes of the initial report, alarm or indication.

• • OR **

(4) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility that requires firefighting support by an offsite fire response agency to extinguish.

Figure 1 ER 1.1 Page 52 Initiating Conditions and Emergency Action Levels Rev. 58 HU4 BASIS INFORMATION HU4 Basis:

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

PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan.

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.

With regard to containment fire 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 reasonably be expected that a fire that burns for 15 minutes would produce sufficient products of combustion to cause fire detectors in multiple zones to alarm.

EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished. 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.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 53 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION HU4 HU4 CONTINUED 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 the Dry Fuel Storage Facility.

EAL #4 If a FIRE within the plant PROTECTED AREA or is of sufficient size to require a response by an offsite firefighting agency, 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 MA9.

Figure 1 ER 1.1 Page 54 Initiating Conditions and Emergency Action Levels Rev. 58 HU7 HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the STED/SED warrant declaration of an Unusual Event.

Operating Mode Applicability: All Emergency Action Levels:

(1) Other conditions exist which in the judgment of the STED/SED 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.

HU7 BASIS INFORMATION HU7 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 STED/SED to fall under the emergency classification level description for a NOUE.

Figure 1 ER 1.1 Page 55 Initiating Conditions and Emergency Action Levels Rev. 58 SYSTEM MALFUNCTION ICS/EALS Recognition Category M Initiating Condition Matrix GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT MG1 Prolonged loss of all MS1 Loss of all offsite and MA1 Loss of all but one MU1 Loss of all offsite AC offsite and all onsite AC all onsite AC power to AC power source to power capability to emergency power to emergency buses. emergency buses for 15 emergency buses for 15 buses for 15 minutes or longer.

Op. Modes: 1, 2, 3, 4 minutes or longer. minutes or longer. Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 MA2 UNPLANNED loss MU2 UNPLANNED loss of Control Room indications of Control Room indications for 15 minutes or longer with a for 15 minutes or longer.

significant transient in Op. Modes: 1, 2, 3, 4 progress.

Op. Modes: 1, 2, 3, 4 MU3 Reactor coolant activity greater than Technical Specification allowable limits.

Op. Modes: 1, 2, 3, 4 MU4 RCS leakage for 15 minutes or longer.

Op. Modes: 1, 2, 3, 4 MS5 Inability to shutdown MA5 Automatic or manual MU5 Automatic or manual the reactor causing a trip fails to shutdown the trip fails to shutdown the challenge to core cooling or reactor and subsequent reactor .

RCS heat removal. manual actions taken at the Op. Modes: 1 Op. Modes: 1 Main Control Board are not successful in shutting down the reactor.

Op. Modes: 1 MU6 Loss of all onsite or offsite communications capabilities.

Op. Modes: 1, 2, 3, 4 MU7 Failure to isolate containment or loss of containment pressure control.

Op. Modes: 1, 2, 3, 4 MG8 Loss of all AC and MS8 Loss of all Vital DC Vital DC power sources for 15 power for 15 minutes or minutes or longer. longer.

Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 MA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

Op. Modes: 1, 2, 3, 4

Figure 1 ER 1.1 Page 56 Initiating Conditions and Emergency Action Levels Rev. 58 MG1 MG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Notes:

• The STED/SED should declare the General Emergency promptly upon determining that 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

(1) a. Loss of ALL offsite and ALL onsite AC power to BOTH AC emergency buses E5 AND E6.

• • AND **

b. ANY of the following:

Restoration of at least one AC emergency bus in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.

Core Cooling (C) CSF RED entry conditions met

Figure 1 ER 1.1 Page 57 Initiating Conditions and Emergency Action Levels Rev. 58 MG1 BASIS INFORMATION MG1 Basis:

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

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1.

This will allow additional time for implementation of offsite protective actions.

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. This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).

The SEPS primary function is to supply power to one 4.16 kV emergency bus, EDE-SWG-5 (E5) or EDE-SWG-6 (E6), in the event of a loss-of-offsite-power (LOOP) and both EDGs fail to start and load.

In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS, CVI, CI, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA) EE-03-007.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 58 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION MG1 MG1 CONTINUED The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover. The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be de-energized before restoring offsite power.

For power restoration from the SEPS, both SEPS diesel generator sets must be functional. Both SEPS engines are required to power the emergency bus and equipment required to maintain core cooling.

The use of the SEPS is recognized in the Emergency Operating Procedures

Reference:

UFSAR Section 8.3.1, AC Power Systems UFSAR Section 8.3.1, AC Power Systems

Figure 1 ER 1.1 Page 59 Initiating Conditions and Emergency Action Levels Rev. 58 MG8 MG8 ECL: General Emergency Initiating Condition: Loss of all AC and Vital DC power sources for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Note:

• The STED/SED should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

(1) a. Loss of ALL offsite and ALL onsite AC power to BOTH AC emergency buses E5 AND E6 for 15 minutes or longer.

• • AND **

b. Indicated voltage is less than 105 V on ALL Vital DC buses 11A, 11B, 11C and 11D for 15 minutes or longer.

Figure 1 ER 1.1 Page 60 Initiating Conditions and Emergency Action Levels Rev. 58 MG8 BASIS INFORMATION MG8 Basis:

This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).

The SEPS primary function is to supply power to one 4.16 kV emergency bus, EDE-SWG-5 (E5) or EDE-SWG-6 (E6), in the event of a loss-of-offsite-power (LOOP) and both EDGs fail to start and load.

In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS, CVI, CI, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA) EE-03-007.

The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover. The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be de-energized before restoring offsite power.

For power restoration from the SEPS, both SEPS diesel generator sets must be functional. Both SEPS engines are required to power the emergency bus and equipment required to maintain core cooling.

The use of the SEPS is recognized in the Emergency Operating Procedures

Reference:

UFSAR Section 8.3.1, AC Power Systems

Figure 1 ER 1.1 Page 61 Initiating Conditions and Emergency Action Levels Rev. 58 MS1 MS1 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Note:

• The STED/SED should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

(1) Loss of ALL offsite and ALL onsite AC power to BOTH AC emergency buses E5 AND E6 for 15 minutes or longer.

Figure 1 ER 1.1 Page 62 Initiating Conditions and Emergency Action Levels Rev. 58 MS1 BASIS INFORMATION MS1 Basis:

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

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

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

This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).

The SEPS primary function is to supply power to one 4.16 kV emergency bus, EDE-SWG-5 (E5) or EDE-SWG-6 (E6), in the event of a loss-of-offsite-power (LOOP) and both EDGs fail to start and load.

In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS, CVI, CI, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA) EE-03-007.

The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover. The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be de-energized before restoring offsite power.

For power restoration from the SEPS, both SEPS diesel generator sets must be functional. Both SEPS engines are required to power the emergency bus and equipment required to maintain core cooling.

The use of the SEPS is recognized in the Emergency Operating Procedures

Reference:

UFSAR Section 8.3.1, AC Power Systems

Figure 1 ER 1.1 Page 63 Initiating Conditions and Emergency Action Levels Rev. 58 MS5 MS5 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.

Operating Mode Applicability: 1 Emergency Action Levels:

(1) a. An automatic or manual trip did not shutdown the reactor.

• • AND **

b. All manual actions to shutdown the reactor have been unsuccessful.

• • AND **

c. EITHER of the following conditions exist:

Core Cooling (C) CSF RED entry conditions met.

Heat Sink (H) CSF RED entry conditions met.

Figure 1 ER 1.1 Page 64 Initiating Conditions and Emergency Action Levels Rev. 58 MS5 BASIS INFORMATION MS5 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator 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.

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

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

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

Figure 1 ER 1.1 Page 65 Initiating Conditions and Emergency Action Levels Rev. 58 MS8 MS8 ECL: Site Area Emergency Initiating Condition: Loss of all Vital DC power for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Note: The STED/SED should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) Indicated voltage is less than 105V on ALL vital DC buses 11A, 11B, 11C and 11D buses for 15 minutes or longer.

MS8 BASIS INFORMATION MS8 Basis:

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 MG8.

Figure 1 ER 1.1 Page 66 Initiating Conditions and Emergency Action Levels Rev. 58 MA1 MA1 ECL: Alert Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Notes:

• The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

1) a AC power capability to BOTH AC emergency buses E5 AND E6 is reduced to a single power source for 15 minutes or longer.

• • AND **

b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

NOTE There are six power sources to consider:

• 345 kV offsite power line 369

• 345 kV offsite power line 363

• 345 kV offsite power line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional

Figure 1 ER 1.1 Page 67 Initiating Conditions and Emergency Action Levels Rev. 58 MA1 BASIS INFORMATION MA1 Basis:

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

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

An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source.

• A loss of all offsite power and loss of all emergency power sources with a single train of emergency buses being back-fed from the unit main generator.

• A loss of emergency power sources with a single train of emergency buses being back-fed from an offsite power source.

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

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

The SEPS primary function is to supply power to one 4.16 kV emergency bus, EDE-SWG-5 (E5) or EDE-SWG-6 (E6), in the event of a loss-of-offsite-power (LOOP) and both EDGs fail to start and load.

In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS, CVI, CI, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA) EE-03-007.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 68 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION MA1 MA1 CONTINUED The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover. The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be de-energized before restoring offsite power.

For power restoration from the SEPS, both SEPS diesel generator sets must be functional. Both SEPS engines are required to power the emergency bus and equipment required to maintain core cooling.

The use of the SEPS is recognized in the Emergency Operating Procedures

Reference:

UFSAR Section 8.3.1, AC Power Systems

Figure 1 ER 1.1 Page 69 Initiating Conditions and Emergency Action Levels Rev. 58 MA2 MA2 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Note: The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.

Reactor Power RCS Level RCS Pressure Core Exit Temperature Level in at least two steam generators Steam Generator Emergency Feed Water Flow

• • AND **

b. ANY of the following transient events in progress.

Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor trip SI actuation

Figure 1 ER 1.1 Page 70 Initiating Conditions and Emergency Action Levels Rev. 58 MA2 BASIS INFORMATION MA2 Basis:

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

This IC addresses the 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 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 cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via ICs FS1 or IC RS1.

Figure 1 ER 1.1 Page 71 Initiating Conditions and Emergency Action Levels Rev. 58 MA5 MA5 ECL: Alert Initiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the Main Control Board are not successful in shutting down the reactor.

Operating Mode Applicability: 1 Emergency Action Level:

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

(1) a. An automatic or manual trip did not shutdown the reactor.

• • AND **

b. Manual actions taken at the MCB are not successful in shutting down the reactor.

Figure 1 ER 1.1 Page 72 Initiating Conditions and Emergency Action Levels Rev. 58 MA5 BASIS INFORMATION MA5 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the MCB 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 MCB since this event entails a significant failure of the RPS.

A manual action at the MCB is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the MCB. 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 MCB.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MS5.

Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC MS5 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.

Figure 1 ER 1.1 Page 73 Initiating Conditions and Emergency Action Levels Rev. 58 MA9 MA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

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

Seismic event (earthquake)

Internal or external flooding event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the STED/SED

• • AND **

b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.

• • OR **

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

Figure 1 ER 1.1 Page 74 Initiating Conditions and Emergency Action Levels Rev. 58 MA9 BASIS INFORMATION MA9 Basis:

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

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

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, 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.

Escalation of the emergency classification level would be via IC FS1 or RS1.

Figure 1 ER 1.1 Page 75 Initiating Conditions and Emergency Action Levels Rev. 58 MU1 MU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Note: The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) Loss of ALL offsite AC power capability to BOTH AC emergency buses E5 AND E6 for 15 minutes or longer.

Figure 1 ER 1.1 Page 76 Initiating Conditions and Emergency Action Levels Rev. 58 MU1 BASIS INFORMATION MU1 Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.

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

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

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

Figure 1 ER 1.1 Page 77 Initiating Conditions and Emergency Action Levels Rev. 58 MU2 MU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels:

Note: The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.

Reactor Power RCS Level RCS Pressure Core Exit Temperature Level in at least two steam generators Steam Generator Emergency Feed Water Flow

Figure 1 ER 1.1 Page 78 Initiating Conditions and Emergency Action Levels Rev. 58 MU2 BASIS INFORMATION MU2 Basis:

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

This IC addresses the 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 reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via IC MA2.

Figure 1 ER 1.1 Page 79 Initiating Conditions and Emergency Action Levels Rev. 58 MU3 MU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2)

(1) RM-6520-1 reading greater than 2,670 mR/hr.

• • OR **

(2) Sample analysis indicates that a reactor coolant activity value is greater than the Limiting Condition for Operation (LCO) specified in Technical Specification 3.4.8 Reactor Coolant System Specific Activity.

Figure 1 ER 1.1 Page 80 Initiating Conditions and Emergency Action Levels Rev. 58 MU3 BASIS INFORMATION MU3 Basis:

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

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

Figure 1 ER 1.1 Page 81 Initiating Conditions and Emergency Action Levels Rev. 58 MU4 MU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2 or 3)

Note: The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) RCS unidentified or PRESSURE BOUNDARY LEAKAGE greater than 10 gpm for 15 minutes or longer.

• • OR **

(2) RCS IDENTIFIED LEAKAGE greater than 25 gpm for 15 minutes or longer.

• • OR **

(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.

Figure 1 ER 1.1 Page 82 Initiating Conditions and Emergency Action Levels Rev. 58 MU4 BASIS INFORMATION MU4 Basis:

IDENTIFIED LEAKAGE

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary to secondary leakage).

PRESSURE BOUNDARY LEAKAGE

a. PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to unidentified leakage",

"pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine). 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, an emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected.

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.

Figure 1 ER 1.1 Page 83 Initiating Conditions and Emergency Action Levels Rev. 58 MU5 MU5 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual trip fails to shutdown the reactor.

Operating Mode Applicability: 1 Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Emergency Action Levels: (1 or 2)

(1) a. An automatic trip did not shutdown the reactor.

• • AND **

b. A subsequent manual action taken at the MCB is successful in shutting down the reactor.

• • OR **

(2) a. A manual trip did not shutdown the reactor.

• • AND **

b. EITHER of the following:

1. A subsequent manual action taken at the MCB is successful in shutting down the reactor.

• • OR **

2. A subsequent automatic trip is successful in shutting down the reactor.

Figure 1 ER 1.1 Page 84 Initiating Conditions and Emergency Action Levels Rev. 58 MU5 BASIS INFORMATION MU5 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the MCB 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 MCB to shutdown the reactor. 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 is unsuccessful, operators will promptly take manual action at another location(s) on the MCB to shutdown the reactor. Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems.

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

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the MCB are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MA5. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC MA5 or FA1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor trip signal be generated as a result of plant work, 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, then this IC and the EALs are not applicable and no classification is warranted.

Figure 1 ER 1.1 Page 85 Initiating Conditions and Emergency Action Levels Rev. 58 MU6 MU6 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2 or 3)

(1) Loss of ALL of the following onsite communication methods:

In-Plant (PBX) Telephones Gai-Tronics Plant Radio System

• • OR **

(2) Loss of ALL of the following ORO communications methods:

Nuclear Alert System (NAS)

Backup NAS Control Room/TSC telephones

• • OR **

(3) Loss of ALL of the following NRC communications methods:

Emergency Notification System (ENS)

Control Room/TSC telephones FTS telephones in the TSC

Figure 1 ER 1.1 Page 86 Initiating Conditions and Emergency Action Levels Rev. 58 MU6 BASIS INFORMATION MU6 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 and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible.

EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the Commonwealth of Massachusetts and State of New Hampshire.

EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Figure 1 ER 1.1 Page 87 Initiating Conditions and Emergency Action Levels Rev. 58 MU7 MU7 ECL: Notification of Unusual Event Initiating Condition: Failure to isolate containment or loss of containment pressure control.

Operating Mode Applicability: 1, 2, 3, 4 Emergency Action Levels: (1 or 2)

(1) a. Failure of containment to isolate when required by an actuation signal.

• • AND **

b. ALL required penetrations are not closed within 15 minutes of the actuation signal.

• • OR **

(2) a. Containment pressure greater than 18 psig.

• • AND **

b. Less than one full train of Containment Building Spray (CBS) is operating per design for 15 minutes or longer.

Figure 1 ER 1.1 Page 88 Initiating Conditions and Emergency Action Levels Rev. 58 MU7 BASIS INFORMATION MU7 Basis:

This IC 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 #1, the containment isolation signal must be generated as the result on an off-normal/accident condition; a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated-should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs.

The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.

The inability to start the required equipment indicates that containment heat removal/depressurization systems 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.

Figure 1 ER 1.1 Page 89 Initiating Conditions and Emergency Action Levels Rev. 58 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Recognition Category C Initiating Condition Matrix GENERAL SITE AREA ALERT UNUSUAL EVENT EMERGENCY EMERGENCY CG1 Loss of reactor CS1 Loss of reactor CA1 Loss of reactor CU1 UNPLANNED vessel/RCS inventory vessel/RCS inventory vessel/RCS inventory. loss of reactor affecting fuel clad affecting core decay Op. Modes: 5, 6 vessel/RCS inventory integrity with heat removal for 15 minutes or containment capability. longer.

challenged. Op. Modes: 5, 6 Op. Modes: 5, 6 Op. Modes: 5, 6 CA2 Loss of all CU2 Loss of all but offsite and all onsite one AC power source AC power to to emergency buses for emergency buses for 15 minutes or longer.

15 minutes or longer. Op. Modes: 5, 6, Op. Modes: 5, 6, Defueled Defueled CA3 Inability to CU3 UNPLANNED maintain the plant in increase in RCS cold shutdown. temperature.

Op. Modes: 5, 6 Op. Modes: 5, 6 CU4 Loss of Vital DC power for 15 minutes or longer.

Op. Modes: 5, 6 CU5 Loss of all onsite or offsite communications capabilities.

Op. Modes: 5, 6, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

Op. Modes: 5, 6

Figure 1 ER 1.1 Page 90 Initiating Conditions and Emergency Action Levels Rev. 58 CG1 CG1 ECL: General Emergency Initiating Condition: Loss of reactor vessel/RCS inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)

Note: The STED/SED should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.

(1) a. RVLIS Full Range < 55% (-141.5 in) for 30 minutes or longer.

• • AND **

b. ANY indication from the Containment Challenge Table C2.

• • OR **

(2) a. Reactor vessel/RCS level cannot be monitored for 30 minutes or longer.

• • AND **

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

RM-6535A-1 (Manipulator Crane) reading greater than 9500 mR/hr RM-6535B-1 (Manipulator Crane) reading greater than 9500 mR/hr Erratic source range monitor indication UNPLANNED increase in Containment Sumps A or B levels of sufficient magnitude to indicate core uncovery.

Visual observation.

• • AND **

c. ANY indication from the Containment Challenge Table C2.

Containment Challenge Table C2 CONTAINMENT INTEGRITY not established

• Containment H2 concentration 6%

UNPLANNED increase in containment pressure

• If CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

Figure 1 ER 1.1 Page 91 Initiating Conditions and Emergency Action Levels Rev. 58 CG1 BASIS INFORMATION CG1 Basis:

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

CONTAINMENT INTEGRITY: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.

With CONTAINMENT INTEGRITY not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit).

A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

Manipulator Crane setpoint of 9500 mR/hr is 95% of the monitor range.

In EAL 2.b, 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.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 92 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION CG1 CG1 CONTINUED The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.

RVLIS LEVEL VESSEL LEVEL

(%) (inches from vessel

~108 119.8 100 81.3 90 31.8 80 -17.7 70 -67.2 63 -101.9 RC-LI-9405, RC-LIT-9467, 60 -116.7 and the Tygon Tube do not 55 -141.5 indicate reactor vessel 50 -166.2 level when actual level is 40 -215.7 less than -95" due to the 30 -265.2 weir on the RCP 20 -314.7 discharge.

10 -364.2 0 -413.7 These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Reference:

FSAR Table 12.3-14

Figure 1 ER 1.1 Page 93 Initiating Conditions and Emergency Action Levels Rev. 58 CS1 CS1 ECL: Site Area Emergency Initiating Condition: Loss of reactor vessel/RCS inventory affecting core decay heat removal capability.

Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2 or 3)

Note: The STED/SED should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.

(1) a. CONTAINMENT INTEGRITY not established.

• • AND **

b. RVLIS Full Range < 63% (-101.9 in) .

• • OR **

(2) a. CONTAINMENT INTEGRITY established.

• • AND **

b. RVLIS Full Range < 55% (-141.5 in).

• • OR **

(3) a. Reactor vessel/RCS level cannot be monitored for 30 minutes or longer.

• • AND **

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

RM-6535A-1 (Manipulator Crane) reading greater than 9500 mR/hr RM-6535B-1 (Manipulator Crane) reading greater than 9500 mR/hr Erratic source range monitor indication UNPLANNED increase in Containment Sumps A or B levels of sufficient magnitude to indicate core uncovery Visual observation.

Figure 1 ER 1.1 Page 94 Initiating Conditions and Emergency Action Levels Rev. 58 CS1 BASIS INFORMATION CS1 Basis:

CONTAINMENT INTEGRITY: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or

2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

This IC addresses a significant and prolonged loss of reactor vessel/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/reactor vessel level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT INTEGRITY 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 INTEGRITY 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.

Manipulator Crane setpoint of 9500 mR/hr is 95% of the monitor range.

The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 95 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION CS1 CS1 CONTINUED RVLIS LEVEL VESSEL LEVEL

(%) (inches from vessel flange)

~108 119.8 100 81.3 90 31.8 80 -17.7 70 -67.2 63 -101.9 RC-LI-9405, RC-LIT-9467, 60 -116.7 and the Tygon Tube do 55 -141.5 not indicate reactor vessel 50 -166.2 level when actual level is 40 -215.7 less than -95" due to the 30 -265.2 weir on the RCP 20 -314.7 discharge.

10 -364.2 0 -413.7 These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or RG1.

Figure 1 ER 1.1 Page 96 Initiating Conditions and Emergency Action Levels Rev. 58 CA1 CA1 ECL: Alert Initiating Condition: Loss of reactor vessel/RCS inventory.

Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)

Note: The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) Loss of reactor vessel/RCS inventory as indicated by RVLIS full range < 64% (-96.9 in.)

• • OR **

(2) a. Reactor vessel/RCS level cannot be monitored for 15 minutes or longer.

• • AND **

b. UNPLANNED increase in Containment Sumps A or B levels due to a loss of reactor vessel/RCS inventory.

Figure 1 ER 1.1 Page 97 Initiating Conditions and Emergency Action Levels Rev. 58 CA1 BASIS INFORMATION CA1 Basis:

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

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For EAL #1, a lowering of water level below 64% indicates that operator actions have not been successful in restoring and maintaining reactor vessel/RCS water level. The heat-up rate of the coolant will 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. 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 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.

Figure 1 ER 1.1 Page 98 Initiating Conditions and Emergency Action Levels Rev. 58 CA2 CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels:

Note:

• The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

(1) Loss of ALL offsite and ALL onsite AC Power to BOTH AC emergency buses E5 AND E6 for 15 minutes or longer.

Figure 1 ER 1.1 Page 99 Initiating Conditions and Emergency Action Levels Rev. 58 CA2 BASIS INFORMATION CA2 Basis:

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.

1-EDE-SWG-5 (E5) and 1-EDE-SWG-6 (E6) are the 4.16 kV emergency buses for Train A and Train B respectively. These buses supply all safety-related loads.

This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).

The SEPS primary function is to supply power to one 4.16 kV emergency bus, EDE-SWG-5 (E5) or EDE-SWG-6 (E6), in the event of a loss-of-offsite-power (LOOP) and both EDGs fail to start and load.

In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS, CVI, CI, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA) EE-03-007.

The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover. The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be de-energized before restoring offsite power.

For power restoration from the SEPS, both SEPS diesel generator sets must be functional. Both SEPS engines are required to power the emergency bus and equipment required to maintain core cooling.

The use of the SEPS is recognized in the Emergency Operating Procedures.

Reference:

UFSAR Section 8.3.1, AC Power Systems Escalation of the emergency classification level would be via IC CS1 or RS1.

Figure 1 ER 1.1 Page 100 Initiating Conditions and Emergency Action Levels Rev. 58 CA3 CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.

Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)

Note: The STED/SED should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

(1) UNPLANNED increase in RCS temperature to greater than 200o F for greater than the duration specified in the following table.

Table C1 - RCS Heat-up Duration Thresholds CONTAINMENT INTEGRITY RCS Status Heat-up Duration Status INTACT and reactor vessel > -

Not applicable 60 minutes*

36 inches Not INTACT or reactor vessel < Established 20 minutes*

-36 inches Not Established 0 minutes

• If RHR is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

• • OR **

(2) UNPLANNED RCS pressure increase greater than 25 psig. (This EAL does not apply during water-solid plant conditions.)

Figure 1 ER 1.1 Page 101 Initiating Conditions and Emergency Action Levels Rev. 58 CA3 BASIS INFORMATION CA3 Basis:

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

INTACT: Capable of being pressurized.

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT INTEGRITY is established but the RCS is not intact, or RCS inventory is reduced. 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 INTEGRITY 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 INTEGRITY is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL #2 provides a pressure-based indication of RCS heat-up. The wide-range RCS pressure transmitters have a range of 0 to 3,000 psig. The main control boards have two post-accident monitoring qualified meters, one for each wide-range RCS pressure transmitter. These meters have major divisions at 100 psig intervals and minor divisions at 50 psig intervals. Since it is possible to read the approximate mid-point between minor divisions, the value is set to 25 psig.

Escalation of the emergency classification level would be via IC CS1 or RS1.

Figure 1 ER 1.1 Page 102 Initiating Conditions and Emergency Action Levels Rev. 58 CA6 CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

Operating Mode Applicability: 5, 6 Emergency Action Levels:

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

Seismic event (earthquake)

Internal or external flooding event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the STED/SED

• • AND **

b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.

• • OR **

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

Figure 1 ER 1.1 Page 103 Initiating Conditions and Emergency Action Levels Rev. 58 CA6 BASIS INFORMATION CA6 Basis:

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

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

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. Systems classified as safety-related.

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.

Escalation of the emergency classification level would be via IC CS1 or RS1.

Figure 1 ER 1.1 Page 104 Initiating Conditions and Emergency Action Levels Rev. 58 CU1 CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of reactor vessel/RCS inventory for 15 minutes or longer.

Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)

Note: The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) UNPLANNED loss of reactor coolant results in reactor vessel/RCS level less than a required lower limit of an operating band, specified by an operating procedure for 15 minutes or longer.

• • OR **

(2) a. Reactor vessel/RCS level cannot be monitored.

• • AND **

b. UNPLANNED increase in Containment Sump A or B level.

Figure 1 ER 1.1 Page 105 Initiating Conditions and Emergency Action Levels Rev. 58 CU1 BASIS INFORMATION CU1 Basis:

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

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor reactor vessel/RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

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 level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL #2 addresses a condition where all means to determine reactor vessel/RCS level have been lost.

In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Figure 1 ER 1.1 Page 106 Initiating Conditions and Emergency Action Levels Rev. 58 CU2 CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.

Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels:

Notes:

• The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For power restoration from the SEPS, both SEPS diesel generator sets must be functional.

(1) a. AC power capability to Both AC emergency buses E5 AND E6 is reduced to a single power source for 15 minutes or longer.

• • AND **

b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

NOTE There are six power sources to consider:

• 345 kV offsite power line 369

• 345 kV offsite power line 363

• 345 kV offsite power line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional

Figure 1 ER 1.1 Page 107 Initiating Conditions and Emergency Action Levels Rev. 58 CU2 BASIS INFORMATION CU2 Basis:

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

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

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus.

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

1-EDE-SWG-5 (E5) and 1-EDE-SWG-6 (E6) are the 4.16 kV emergency buses for Train A and Train B respectively. These buses supply all safety-related loads.

This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).

The SEPS primary function is to supply power to one 4.16 kV emergency bus, EDE-SWG-5 (E5) or EDE-SWG-6 (E6), in the event of a loss-of-offsite-power (LOOP) and both EDGs fail to start and load.

In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS, CVI, CI, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA) EE-03-007.

The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover. The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be de-energized before restoring offsite power.

[BASIS CONTINUED ON NEXT PAGE]

Figure 1 ER 1.1 Page 108 Initiating Conditions and Emergency Action Levels Rev. 58 BASIS INFORMATION CU2 CU2 CONTINUED For power restoration from the SEPS, both SEPS diesel generator sets must be functional. Both SEPS engines are required to power the emergency bus and equipment required to maintain core cooling.

The use of the SEPS is recognized in the Emergency Operating Procedures.

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

Reference:

UFSAR Section 8.3.1, AC Power Systems

Figure 1 ER 1.1 Page 109 Initiating Conditions and Emergency Action Levels Rev. 58 CU3 CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.

Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)

Note: The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) UNPLANNED increase in RCS temperature to greater than 200o F.

• • OR **

Loss of ALL RCS temperature and reactor vessel/RCS level indication for 15 minutes or longer.

Figure 1 ER 1.1 Page 110 Initiating Conditions and Emergency Action Levels Rev. 58 CU3 BASIS INFORMATION CU3 Basis:

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

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT INTEGRITY is not established during this event, the STED/SED should also refer to IC CA3.

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.

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.

Figure 1 ER 1.1 Page 111 Initiating Conditions and Emergency Action Levels Rev. 58 CU4 CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.

Operating Mode Applicability: 5, 6 Emergency Action Levels:

Note: The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) Indicated voltage is less than 105V on required Vital DC buses associated with the Protected Train for 15 minutes or longer.

Train A 11A and 11C Train B 11B and 11D

Figure 1 ER 1.1 Page 112 Initiating Conditions and Emergency Action Levels Rev. 58 CU4 BASIS INFORMATION CU4 Basis:

This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service.

Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, required means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable),

then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

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

Per DBD-ED-05, the DC bus voltage range within which the 125 Volt DC system is considered operable is 105 volts minimum to 140 volts maximum. The vital DC Buses (Switchgear) are SWG-11A and 11C for Train A and SWG-11B and 11D for Train B.

Reference:

UFSAR Section 8.3.2, DC Power System Procedure OS1248.01, Loss of a Vital 125 VDC Bus Procedure VPRO F5278, Loss of All Vital DC Power DBD-ED-05, 125 VDC System Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.

Figure 1 ER 1.1 Page 113 Initiating Conditions and Emergency Action Levels Rev. 58 CU5 CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: (1 or 2 or 3)

(1) Loss of ALL of the following onsite communication methods:

In-Plant (PBX) Telephones Gai-Tronics Plant Radio System

• • OR **

(2) Loss of ALL of the following ORO communications methods:

Nuclear Alert System (NAS)

Backup NAS Control Room/TSC telephones

• • OR **

(3) Loss of ALL of the following NRC communications methods:

Emergency Notification System (ENS)

Control Room/TSC telephones FTS telephones in the TSC

Figure 1 ER 1.1 Page 114 Initiating Conditions and Emergency Action Levels Rev. 58 CU5 BASIS INFORMATION CU5 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 and the NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible.

EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are Commonwealth of Massachusetts and State of New Hampshire.

EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Figure 1 ER 1.1 Page 115 Initiating Conditions and Emergency Action Levels Rev. 58 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS Recognition Category E Initiating Condition Matrix UNUSUAL EVENT EU1 Damage to a loaded cask CONFINEMENT BOUNDARY.

Op. Modes: All

Figure 1 ER 1.1 Page 116 Initiating Conditions and Emergency Action Levels Rev. 58 EU1 EU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: All Emergency Action Levels:

Note:

The on-contact dose rate may be determined based on measurement of a dose rate at some distance from the cask (1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by ANY of the following on-contact surface radiation readings greater than:

1600 mrem/hr at the front bird screen 4 mrem/hr at the door centerline 4 mrem/hr at the end shield wall exterior

Figure 1 ER 1.1 Page 117 Initiating Conditions and Emergency Action Levels Rev. 58 EU1 BASIS INFORMATION EU1 Basis:

CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage.

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed in the Horizontal Storage Module. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of damage is determined by radiological survey. The technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HA1.

Reference:

Appendix A to Certificate Of Compliance No. 1030 NUHOMS HD System Generic Technical Specifications 5.4.3.

Figure 1 ER 1.1 Page 118 Initiating Conditions and Emergency Action Levels Rev. 58 FISSION PRODUCT BARRIER ICS/EALS Recognition Category F Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier.

FG1 Op. Modes: 1, 3, 2, 4 SITE AREA EMERGENCY Loss or Potential Loss of any two barriers.

FS1 Op. Modes: 1, 3, 2, 4 ALERT Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.

FA1 Op. Modes: 1, 3, 2, 4

Figure 1 ER 1.1 Page 119 Initiating Conditions and Emergency Action Levels Rev. 58 Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FG1 GENERAL EMERGENCY FS1 SITE AREA EMERGENCY FA1 ALERT Loss of any two barriers and Loss or Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either Potential Loss of the third barrier. the Fuel Clad or RCS barrier.

Figure 1 ER 1.1 Page 120 Initiating Conditions and Emergency Action Levels Rev. 58 Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage Not Applicable A. Core Cooling (C) A. An automatic or A. Operation of a second A. A leaking or Not Applicable CSF - ORANGE manual SI actuation is charging pump in the RUPTURED SG is entry conditions met required by EITHER normal charging FAULTED outside of (NOTE 1) of the following: mode is required by containment.

1. UNISOLABLE EITHER of the RCS leakage following:

OR 1. UNISOLABLE RCS leakage

2. SG tube RUPTURE. OR

2. SG tube leakage.

OR B. RCS Integrity (P)

CSF - RED entry conditions met with RCS press > 300 psig. (NOTE 1).

2. Inadequate Heat Removal 2. Inadequate Heat Removal 2. Inadequate Heat Removal A. Core Cooling (C) A. Core Cooling (C) Not Applicable A. Heat Sink (H) CSF - Not Applicable A. Core Cooling (C) CSF CSF - RED entry CSF - ORANGE RED entry conditions - RED entry conditions conditions met. entry conditions met. met. (NOTE 1) met for 15 minutes or (NOTE 1) (NOTE 1) longer. (NOTE 1)

OR B. Heat Sink (H) CSF -

RED entry conditions met. (NOTE 1)

Figure 1 ER 1.1 Page 121 Initiating Conditions and Emergency Action Levels Rev. 58

3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation A. Post LOCA Not Applicable A. Post LOCA Radiation Not Applicable Not Applicable A. Post LOCA Radiation Radiation Monitors Monitors Monitors RM 6576A-1 or RM RM 6576A-1 or RM RM 6576A-1 or RM 6576B-1 6576B-1 6576B-1 95 R/hr. 16 R/hr. 1,305 R/hr.

OR B. RCS activity > 300 uCi/gm Dose Equivalent I 131 as determined per Procedure CS0925.01, Reactor Coolant Post Accident Sampling.

4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not Applicable Not Applicable Not Applicable Not Applicable A. Containment isolation A. Containment (Z) CSF -

is required RED entry conditions AND met. (NOTE 1)

EITHER of the OR following: B. Containment H2

1. Containment concentration 6%

integrity has been OR lost based on STED/SED C. 1. Containment judgment. pressure > 18 psig OR AND

2. UNISOLABLE 2. Less than one full pathway from the train of containment to Containment the environment Building Spray exists. (CBS) is operating per OR design for 15 B. Indications of RCS minutes or longer.

leakage outside of containment.

Figure 1 ER 1.1 Page 122 Initiating Conditions and Emergency Action Levels Rev. 58

5. STED/SED Judgment 5. STED/SED Judgment 5. STED/SED Judgment A. ANY condition in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the the opinion of the opinion of the opinion of the opinion of the opinion of the opinion of the STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that indicates Loss of the indicates Potential indicates Loss of the indicates Potential indicates Loss of the indicates Potential Loss Fuel Clad Barrier. Loss of the Fuel Clad RCS Barrier. Loss of the RCS Containment Barrier. of the Containment Barrier. Barrier. Barrier.

NOTE 1: Refer to ER 1.1, Section 1.1, Discussion concerning the proper use of CSFSTs as EALs

Figure 1 ER 1.1 Page 123 Initiating Conditions and Emergency Action Levels Rev. 58 Basis Information For Fission Product Barrier Table FUEL CLAD BARRIER THRESHOLDS:

The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

1. RCS or SG Tube Leakage There is no Loss threshold associated with RCS or SG Tube Leakage.

Potential Loss 1.A This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

2. Inadequate Heat Removal Loss 2.A This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

Potential Loss 2.A This reading indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.

Potential Loss 2.B This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; 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.A; 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.

As a potential loss indication, developers should consider including a threshold the same as, or similar to, Core Cooling Orange entry conditions met in accordance with the guidance at the front of this section.

As a potential loss indication, developers should consider including a threshold the same as, or similar to, Heat Sink Red entry conditions met in accordance with the guidance at the front of this section.

Figure 1 ER 1.1 Page 124 Initiating Conditions and Emergency Action Levels Rev. 58

3. RCS Activity / Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 uCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 uCi/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.

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

4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)

5. STED/SED Judgment Loss 5.A This threshold addresses any other factors that may be used by the STED/SED in determining whether the Fuel Clad Barrier is lost.

Potential Loss 5.A This threshold addresses any other factors that may be used by the STED/SED in determining whether the Fuel Clad Barrier is potentially lost. The STED/SED should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Figure 1 ER 1.1 Page 125 Initiating Conditions and Emergency Action Levels Rev. 58 Before invoking this judgment (discretionary) EAL, the STED/SED should consider the following information. The values are plotted roughly to scale. Use of this judgment EAL requires a totality of indications that lead the STED or SED to conclude that RCS radioactivity is near (approaching), equal to or greater than 300 uCi/cc Dose Equivalent Iodine-131.

Fuel Clad Barrier Loss value IC FA1 = 300 uCi/gm DE I-131 based on PASS port dose rate (per CS0925.01)

IC MU3 When considering a classification in this range, the STED or SED will need to conclude that RCS radioactivity is near (approaching),

equal to or greater than 300 uCi/gm Dose Equivalent Iodine-131 before declaring the fuel clad barrier lost or potentially lost.

Letdown monitor off-scale high value = 10,000 mR/hr IC MU3, EAL #1: letdown Typical letdown monitor reading monitor value = 2,670 mR/hr.

15 to 20 mR/hr. May go up to This corresponds to RCS activity

~250 mR/hr during a crud burst. at the Technical Specification allowable limit.

Figure 1 ER 1.1 Page 126 Initiating Conditions and Emergency Action Levels Rev. 58 RCS BARRIER THRESHOLDS:

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.

1. RCS or SG Tube Leakage Loss 1.A 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.A will also be met.

Potential Loss 1.A 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) 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 level.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met.

Potential Loss 1.B 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).

Figure 1 ER 1.1 Page 127 Initiating Conditions and Emergency Action Levels Rev. 58

2. Inadequate Heat Removal There is no Loss threshold associated with Inadequate Heat Removal.

Potential Loss 2.A This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using 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.B; 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.

3. RCS Activity / Containment Radiation Loss 3.A 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. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only.

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

4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)

5. STED/SED Judgment Loss 5.A This threshold addresses any other factors that may be used by the STED/SED in determining whether the RCS Barrier is lost.

Potential Loss 5.A This threshold addresses any other factors that may be used by the STED/SED in determining whether the RCS Barrier is potentially lost. The STED/SED should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Figure 1 ER 1.1 Page 128 Initiating Conditions and Emergency Action Levels Rev. 58 CONTAINMENT BARRIER THRESHOLDS:

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 ECL from Alert to a Site Area Emergency or a General Emergency.

1. RCS or SG Tube Leakage Loss 1.A 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 and Loss 1.A, 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 MU3 for the fuel clad barrier (i.e., RCS activity values) and IC MU4 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump.

These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition).

The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve do meet this threshold.

Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component. These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

Figure 1 ER 1.1 Page 129 Initiating Conditions and Emergency Action Levels Rev. 58 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 Unusual Event per MU4 MU4 Requires operation of a second Site Area Emergency charging pump (RCS Barrier Alert per FA1 per FS1 Potential Loss)

Requires an automatic or Site Area Emergency manual SI actuation (RCS Alert per FA1 per FS1 Barrier Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

2. Inadequate Heat Removal There is no Loss threshold associated with Inadequate Heat Removal.

Potential Loss 2.A 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 STED/SED should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

Severe accident analyses 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.

Figure 1 ER 1.1 Page 130 Initiating Conditions and Emergency Action Levels Rev. 58

3. RCS Activity / Containment Radiation There is no Loss threshold associated with RCS Activity / Containment Radiation.

Potential Loss 3.A 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.

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.

4. Containment Integrity or Bypass Loss 4.A 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 thresholds 4.A.1 and 4.A.2.

4.A.1 - 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 STED/SED will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

4.A.2 - 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 Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Figure 1 ER 1.1 Page 131 Initiating Conditions and Emergency Action Levels Rev. 58 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.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 1.A.

Loss 4.B 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.

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.A to be met.

Potential Loss 4.A If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.

Figure 1 ER 1.1 Page 132 Initiating Conditions and Emergency Action Levels Rev. 58 Potential Loss 4.B The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.

Potential Loss 4.C 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. 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 are either lost or performing in a degraded manner.

5. STED/SED Judgment Loss 5.A This threshold addresses any other factors that may be used by the STED/SED in determining whether the Containment Barrier is lost.

Potential Loss 5.A This threshold addresses any other factors that may be used by the STED/SED in determining whether the Containment Barrier is potentially lost. The STED/SED should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

ER 1.1 Page 133 Rev. 58 Figure 2 Definitions (Sheet 1 of 3)

CONFINEMENT BOUNDARY: - The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage.

CONTAINMENT INTEGRITY:- The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Emergency Action Level (EAL): A pre-determined, 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: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.

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, observable threshold indicating the loss or potential loss of a fission product barrier.

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

ER 1.1 Page 134 Rev. 58 Figure 2 Definitions (Sheet 2 of 3)

HOSTILE ACTION: An act toward a NPP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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: Entry into an area requires extraordinary measures to facilitate entry of personnel into the affected room/area by installing temporary shielding, requiring use of non-routine protective equipment, or requesting an extension in dose limits beyond normal administrative limits.

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. (Dry Fuel Storage Facility)

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.

INTACT: Capable of being pressurized.

OWNER CONTROLLED AREA: The site property owned by, or otherwise under the control of, the licensee.

PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan.

REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal.

RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

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.

ER 1.1 Page 135 Rev. 58 Figure 2 Definitions (Sheet 2 of 3)

SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

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

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or

2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

ER 1.1 Page 136 Rev. 58 Figure 3 Acronyms and Abbreviations (Sheet 1 of 2)

AC ...................................................................................................................... Alternating Current AOP................................................................................................. Abnormal Operating Procedure ATWS ................................................................................... Anticipated Transient Without Scram CDE...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CTMT/CNMT ............................................................................................................... Containment CSF ............................................................................................................. Critical Safety Function CSFST ...................................................................................... Critical Safety Function Status Tree DBA .............................................................................................................. Design Basis Accident DC .............................................................................................................................. Direct Current EAL ........................................................................................................... Emergency Action Level ECCS............................................................................................ Emergency Core Cooling System ECL ................................................................................................ Emergency Classification Level EOF ..................................................................................................Emergency Operations Facility EOP ............................................................................................... Emergency Operating Procedure EPA ............................................................................................. Environmental Protection Agency FEMA ............................................................................. Federal Emergency Management Agency FSAR................................................................................................... Final Safety Analysis Report GE ...................................................................................................................... General Emergency IC........................................................................................................................ Initiating Condition ID ............................................................................................................................. Inside Diameter ISFSI ............................... Independent Spent Fuel Storage Installation (Dry Fuel Storage Facility)

Keff .................................................................................... Effective Neutron Multiplication Factor LCO............................................................................................... Limiting Condition of Operation LOCA ........................................................................................................Loss of Coolant Accident MCB .................................................................................................................. Main Control Board MSIV..................................................................................................... Main Steam Isolation Valve MSL ....................................................................................................................... Main Steam Line mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man MW ....................................................................................................................................Megawatt NEI ............................................................................................................. Nuclear Energy Institute NPP .................................................................................................................. Nuclear Power Plant NRC .............................................................................................. Nuclear Regulatory Commission NSSS ................................................................................................. Nuclear Steam Supply System NORAD ................................................................. North American Aerospace Defense Command OBE....................................................................................................... Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODCM........................................................................................... Offsite Dose Calculation Manual ORO ................................................................................................ Off-site Response Organization PA .............................................................................................................................. Protected Area PAG....................................................................................................... Protective Action Guideline PRA ................................................................................................... Probabilistic Risk Assessment PWR ........................................................................................................ Pressurized Water Reactor

ER 1.1 Page 137 Rev. 58 Figure 3 Acronyms and Abbreviations (Sheet 2 of 2)

PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ......................................................................................Roentgen Equivalent Man RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RVLIS ...................................................................... Reactor Vessel Level Instrumentation System SAR .............................................................................................................. Safety Analysis Report SAS ........................................................................................................... Secondary Alarm Station SBO ......................................................................................................................... Station Blackout SCBA ..................................................................................... Self-Contained Breathing Apparatus SG ...........................................................................................................................Steam Generator SI .............................................................................................................................. Safety Injection SPDS ............................................................................................ Safety Parameter Display System SRO ............................................................................................................ Senior Reactor Operator TEDE ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel TSC .......................................................................................................... Technical Support Center WOG .................................................................................................. Westinghouse Owners Group

ER 1.1 Page 138 Rev. 58 Figure 4 Additional Basis Information (Sheet 1 of 10) 1 REGULATORY BACKGROUND 1.1 OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.

Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology of this document.

• 10 CFR § 50.47(a)(1)(i)

• 10 CFR § 50.47(b)(4)

• 10 CFR § 50.54(q)

• 10 CFR § 50.72(a)

• 10 CFR § 50, Appendix E, IV.B, Assessment Actions

• 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents. Three documents of particular relevance to NEI 99-01 are:

• NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]

• NUREG-1022, Event Reporting Guidelines 10 CFR § 50.72 and § 50.73

• Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR § 50.47 emergency plan.

ISFSI ICs/EALs. IC EU1 covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design.

ER 1.1 Page 139 Rev. 58 Figure 4 Additional Basis Information (Sheet 2 of 10)

The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees. NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.

1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plants design basis and flooded the sites emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling.

Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.

NRC Order EA-12-051 states, in part, All licensees shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. To this end, all licensees must provide:

• A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;

• A display in an area accessible following a severe event; and

• Independent electrical power to each instrument channel and provide an alternate remote power connection capability.

ER 1.1 Page 140 Rev. 58 Figure 4 Additional Basis Information (Sheet 3 of 10)

NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, provides guidance for complying with NRC Order EA-12-051.

NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within existing IC RA2, and new ICs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.

It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.

The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). In accordance with this regulation, licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the effectiveness of the plan. As a result of the licensee's determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.

1.4 ORGANIZATION AND PRESENTATION OF INFORMATION The schemes information is organized by Recognition Category in the following order.

R - Abnormal Radiation Levels / Radiological Effluent C - Cold Shutdown / Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)

F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety M - System Malfunction

ER 1.1 Page 141 Rev. 58 Figure 4 Additional Basis Information (Sheet 4 of 10) 1.5 IC AND EAL MODE APPLICABILITY The following table shows which Recognition Categories are applicable in each plant mode.

The ICs and EALs for a given Recognition Category are applicable in the indicated modes.

MODE APPLICABILITY MATRIX Category Mode R C E F H M Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X Operating Modes Technical Specifications TABLE 1.2 MODE Reactivity  % Rated Thermal Average Coolant Condition, Keff Power* Temperature

1. Power Operation > 0.99 > 5% 350°F

2. Startup 0.99 < 5% 350°F

3. Hot Standby < 0.99 0 350°F

4. Hot Shutdown < 0.99 0 350 °F > Tavg >200 °F

5. Cold Shutdown < 0.99 0 < 200 °F

6. Refueling** NA 0 < 140 °F NA Defueled All fuel removed from the reactor vessel (full core offload during refueling or extended outage)

• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

ER 1.1 Page 142 Rev. 58 Figure 4 Additional Basis Information (Sheet 5 of 10) 1.6 BASIS DOCUMENT The basis document is an integral part of an emergency classification scheme. The material in this document supports proper emergency classification decision-making by providing informing background and development information in a readily accessible format. It can be referred to in training situations and when making an actual emergency classification, if necessary. The document is also useful for establishing configuration management controls for EP-related equipment and explaining an emergency classification to offsite authorities.

The content of the basis document includes:

• A site-specific Mode Applicability Matrix and description of operating modes.

• A discussion of the emergency classification and declaration process

• Each Initiating Condition along with the associated EALs or fission product barrier thresholds, Operating Mode Applicability, Notes and Basis information

• A listing of acronyms and defined terms A basis section should not contain information that could modify the meaning or intent of the associated IC or EAL. Such information should be incorporated within the IC or EAL statements, or as an EAL Note. Information in the Basis should only clarify and inform decision-making for an emergency classification.

Basis information should be readily available to be referenced, if necessary, by the Short Term Emergency Director/Site Emergency Director (STED/SED). For example, a copy of the basis document could be maintained in the appropriate emergency response facilities.

Because the information in a basis document can affect emergency classification decision-making (e.g., the STED/SED 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).

1.7 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA The criteria/values used in several EALs and fission product barrier thresholds may be drawn from AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.

ER 1.1 Page 143 Rev. 58 Figure 4 Additional Basis Information (Sheet 6 of 10) 2 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1 GENERAL CONSIDERATIONS When making an emergency classification, the STED/SED 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 Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.

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.

All emergency classification assessments should 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, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration.

For ICs and EALs that have a stipulated time duration, the STED/SED 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.

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 10 § CFR 50.72.

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; the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15- minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e.,

this is the time that the EAL information is first available).

ER 1.1 Page 144 Rev. 58 Figure 4 Additional Basis Information (Sheet 7 of 10)

The NRC expects licensees to establish the capability to initiate and complete EAL- related analyses within a reasonable period of time.

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 scheme provides the STED/SED 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 STED/SED 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 into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

2.2 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(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures.

When assessing an EAL that specifies a time duration for the off-normal 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.

2.3 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, 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, 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.

ER 1.1 Page 145 Rev. 58 Figure 4 Additional Basis Information (Sheet 8 of 10) 2.4 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.

2.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the STED/SED 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 STED/SED, 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.

2.6 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.

ER 1.1 Page 146 Rev. 58 Figure 4 Additional Basis Information (Sheet 9 of 10)

The following approach to downgrading or terminating an ECL is recommended.

ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with plant procedures.

Alert Downgrade or terminate the emergency in accordance with plant procedures.

Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures.

Site Area Emergency with Terminate the emergency and enter recovery in long-term plant damage accordance with plant procedures.

General Emergency Terminate the emergency and enter recovery in accordance with plant procedures.

2.7 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 EAL must be declared regardless of its continued presence at the time of declaration. Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.

2.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the ICs and/or EALs contained in this document 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. The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response - In instances where 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.

ER 1.1 Page 147 Rev. 58 Figure 4 Additional Basis Information (Sheet 10 of 10)

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 auxiliary feedwater system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal 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 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 where an operator is able to take a successful corrective action prior to the STED/SED completing the review 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.

2.9 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 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 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.

ER 1.1 Page 148 Rev. 58 Figure 5 Summary of Changes (Sheet 1 of 1)

Rev. 58: (PCR 02205495 July 2017)

Complete rewrite to new EAL scheme based on NEI 99-01 rev. 06 per LAR 15-02 (AR 02101091).

Rev. 57: (PCR 02182157 February 2017)

Replaced Reference to ER 5.7 with ER 5.3 as there is now only one standard procedure for all Raddose-V users.

Rev. 56: (PCR 02123671)

Removed paragraph from discussion section regarding the Site Emergency Director obtaining concurrence from the EOF Coordinator prior to reclassifying category 'A' EALs (AR 2109262).

This actionable step has been moved to the SEDs checklist in ER 3.1. Moved actionable steps from discussion section to section 5.1 (AR 2109262). Changed reference for reporting procedure from the Regulatory Compliance Manual to LI-AA-102-1001, Regulatory Reporting.

HU1 - Changed yellow "EVENT" light to red "EVENT" light to support new seismic monitoring system (EC 282184).

HA1 - Changed yellow "EVENT" light to red "EVENT" light and red "OBE" light to yellow "OBE" light support new seismic monitoring system (EC 282184).

Rev. 55: (PCR 2007831)

Added clarification to discussion section to raise awareness that certain plant conditions may mask EAL indications. (AR 1986114).

Rev. 54:

Added reactor vessel level note to cold shutdown EALs with table that compares RVLIS indication to vessel level with the vessel flange as the datum (AR 01948034).

In SA5 note box, replaced 345 KV line names with 345 KV line numbers (AR 01930975).

EMERGENCY INITIATING CONDITION MATRIX Modes 1, 2, 3, and 4 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Category R- Abnormal Rad Levels/Radiological Effluent RG1 Release of gaseous radioactivity RS1 Release of gaseous radioactivity RA1 Release of gaseous or liquid RU1 Release of gaseous or liquid resulting in offsite dose > 1,000 resulting in offsite dose > 100 mrem radioactivity resulting in offsite dose radioactivity > 2 times the ODCM mrem TEDE or 5,000 mrem thyroid TEDE or 500 mrem thyroid CDE > 10 mrem TEDE or 50 mrem thyroid limits for 60 minutes CDE Op. Modes: All CDE Op. Modes: All Op. Modes: All Op. Modes: All RG2 Spent fuel pool level cannot be RS2 Spent fuel pool level at 1.5 ft. (Level 3) RA2 Significant lowering of water level RU2 UNPLANNED loss of water level restored to at least 1.5 ft. (Level 3) Op. Modes: All above, or damage to, irradiated fuel. above irradiated fuel for 60 minutes or longer. Op. Modes: All Op. Modes: All Op. Modes: All RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.

Op. Modes: All Category E - Events Related to ISFSI Malfunction EU1 Damage to a loaded cask CONFINEMENT BOUNDARY Op. Mode: All Category H - Hazards and Other Conditions Affecting Plant Safety HS1 HOSTILE ACTION within the HA1 HOSTILE ACTION within the HU1 Confirmed SECURITY CONDITION PROTECTED AREA. OWNER CONTROLLED AREA or or threat.

Op. Modes: All airborne attack threat within 30 Op. Modes: All minutes Op. Modes: All HU2 Seismic event greater than OBE levels.

Op. Modes: All HU3 Hazardous event.

Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.

Op. Modes: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, shutdown or cooldown Op. Modes: All HS6 Inability to control a key safety function HA6 Control Room evacuation resulting in from outside the Control Room transfer of plant control to alternate Op. Modes: All locations Op. Modes: All HG7 Other conditions exist which in the HS7 Other conditions exist which in the HA7 Other conditions exist which in the HU7 Other conditions exist which in the judgment of the STED/SED warrant judgment of the STED/SED warrant judgment of the STED/SED warrant judgment of the STED/SED warrant declaration of a General Emergency declaration of a Site Area Emergency declaration of an Alert declaration of an Unusual Event Op. Modes: All Op. Modes: All Op. Modes: All Op. Modes: All Category M - System Malfunction MG1 Prolonged loss of all offsite and all MS1 Loss of all offsite and all onsite AC MA1 Loss of all but one AC power source MU1 Loss of all offsite AC power onsite AC power to emergency power to emergency buses for 15 to emergency buses for 15 minutes capability to emergency buses for buses minutes or longer or longer 15 minutes or longer AND Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4

• Restoration of at least one AC emergency bus in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely.

OR

• Core Cooling (C) CSF RED entry conditions met Op. Modes: 1, 2, 3, 4 MA2 UNPLANNED loss of Control Room MU2 UNPLANNED loss of Control Room indications for 15 minutes or longer indications for 15 minutes or longer with a significant transient in Op. Modes: 1, 2, 3, 4 progress.

Op. Modes: 1, 2, 3, 4 MU3 Reactor coolant activity greater than Technical Specification allowable limits Op. Modes: 1, 2, 3, 4 MU4 RCS leakage for 15 minutes or longer Op. Modes: 1, 2, 3, 4 MS5 Inability to shutdown the reactor MA5 Automatic or manual trip fails to MU5 Automatic or manual trip fails to causing a challenge to core cooling or shutdown the reactor, and shutdown the reactor RCS heat removal subsequent manual actions taken at Op. Modes: 1 Op. Modes: 1 the Main Control Board are not successful in shutting down the reactor Op. Modes: 1 MU6 Loss of all onsite or offsite communications capabilities Op. Modes: 1, 2, 3, 4 MU7 Failure to isolate containment or loss of containment pressure control Op Modes: 1, 2, 3, 4 MG8 Loss of all AC and Vital DC power MS8 Loss of all Vital DC power for sources for 15 minutes or longer 15 minutes or longer Op. Modes: 1, 2, 3, 4 Op. Modes: 1, 2, 3, 4 MA9 Hazardous event affecting a SAFETY SYSTEM needed for the current Op. Modes: 1, 2, 3, 4 Modes 1, 2, 3 and 4 ER 1.1A Rev. 47

EMERGENCY INITIATING CONDITION MATRIX Modes 5, 6, and Defueled GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Category A - Abnormal Rad Levels/Radiological Effluent RG1 Release of gaseous radioactivity RS1 Release of gaseous radioactivity RA1 Release of gaseous or liquid RU1 Release of gaseous or liquid resulting in offsite dose > 1,000 resulting in offsite dose > 100 mrem radioactivity resulting in offsite dose radioactivity > 2 times the ODCM mrem TEDE or 5,000 mrem thyroid TEDE or 500 mrem thyroid CDE > 10 mrem TEDE or 50 mrem thyroid limits for 60 minutes CDE Op. Modes: All CDE Op. Modes: All Op. Modes: All Op. Modes: All RG2 Spent fuel pool level cannot be RS2 Spent fuel pool level at 1.5 ft. (Level 3) RA2 Significant lowering of water level RU2 UNPLANNED loss of water level restored to at least 1.5 ft. (Level 3) Op. Modes: All above, or damage to, irradiated fuel. above irradiated fuel for 60 minutes or longer. Op. Modes: All Op. Modes: All Op. Modes: All RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.

Op. Modes: All Category E - Events Related to ISFSI Malfunction EU1 Damage to a loaded cask CONFINEMENT BOUNDARY Op. Mode: All Category H - Hazards and Other Conditions Affecting Plant Safety HS1 HOSTILE ACTION within the HA1 HOSTILE ACTION within the HU1 Confirmed SECURITY CONDITION PROTECTED AREA OWNER CONTROLLED AREA or or threat Op. Modes: All airborne attack threat within 30 Op. Modes: All minutes Op. Modes: All HU2 Seismic event greater than OBE levels.

Op. Modes: All HU3 Hazardous event.

Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.

Op. Modes: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, shutdown or cooldown Op. Modes: All HS6 Inability to control a key safety function HA6 Control Room evacuation resulting in from outside the Control Room transfer of plant control to alternate Op. Modes: All locations Op. Modes: All HG7 Other conditions exist which in the HS7 Other conditions exist which in the HA7 Other conditions exist which in the HU7 Other conditions exist which in the judgment of the STED/SED warrant judgment of the STED/SED warrant judgment of the STED/SED warrant judgment of the STED/SED warrant declaration of a General Emergency declaration of a Site Area Emergency declaration of an Alert declaration of an Unusual Event Op. Modes: All Op. Modes: All Op. Modes: All Op. Modes: All Category C - Cold Shutdown/Refueling System Malfunction CG1 Loss of reactor vessel/RCS inventory CS1 Loss of reactor vessel/RCS inventory CA1 Loss of reactor vessel/RCS inventory CU1 UNPLANNED loss of reactor affecting fuel clad integrity with affecting core decay heat removal Op. Modes: 5, 6 vessel/RCS inventory for 15 containment challenged capability minutes or longer Op. Modes: 5, 6 Op. Modes: 5, 6 Op. Modes: 5, 6 CA2 Loss of all offsite and all onsite AC CU2 Loss of all but one AC power power to emergency buses for 15 source to emergency buses for 15 minutes or longer minutes or longer Op. Modes: 5, 6, Defueled Op. Modes: 5, 6, Defueled CA3 Inability to maintain the plant in cold CU3 UNPLANNED increase in RCS shutdown temperature.

Op. Modes: 5, 6 OR Loss of ALL RCS temperature and reactor vessel/RCS level indication for 15 minutes or longer Op. Modes: 5, 6 CU4 Loss of Vital DC power for 15 minutes or longer Op. Modes: 5, 6 CU5 Loss of all onsite or offsite communications capabilities Op. Modes: 5, 6, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode Op. Modes: 5, 6 Modes 5, 6, and Defueled ER 1.1B Rev. 47

FISSION PRODUCT BARRIER DEGRADATION MATRIX Modes 1, 2, 3, and 4 Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier Sub-Category Loss Potential Loss Loss Potential Loss Loss Potential Loss Core Cooling (C) RED entry conditions met for Core Cooling (C) ORANGE entry conditions RCS Integrity (P) RED entry conditions met 15 minutes or longer met with RCS press > 300 psig.

Core Cooling (C) RED entry conditions met. OR

1. CSF Status OR OR (Note 1) Containment (Z) CSF - RED entry conditions Heat Sink (H) RED entry conditions met. Heat Sink (H) RED entry conditions met.

met.

(Note 1) (Note 1)

(Note 1)

RCS activity > 300 uCi/gm Dose Equivalent I-131 (as determined per Procedure

2. RCS Activity CS0925.01, Reactor Coolant Post Accident Sampling)

Operation of a second charging pump in the An automatic or manual SI actuation is normal charging mode is required by EITHER required by EITHER of the following:

of the following:

1. UNISOLABLE RCS leakage Indications of RCS leakage outside of

3. RCS Leakage 1. UNISOLABLE RCS leakage containment.

OR OR

2. SG tube RUPTURE

2. SG tube leakage.

A leaking or RUPTURED SG is FAULTED

4. S/G Rupture or Fault outside of containment.

Containment isolation is required AND Containment H2 concentration 6%

EITHER of the following: OR

1. Containment integrity has been lost 1. Containment pressure > 18 psig

5. Containment Integrity based on STED/SED judgment.

AND OR

2. Less than one full train of Containment

2. UNISOLABLE pathway from the Building Spray (CBS) is operating per containment to the environment exists. design for 15 minutes or longer

6. Containment Post-LOCA Radiation Monitors Post-LOCA Radiation Monitors Post-LOCA Radiation Monitors Radiation Monitor RM-6576A-1 or RM-6576B-1 95 R/hr RM-6576A-1 or RM-6576B-1 16 R/hr RM-6576A-1 or RM-6576B-1 1,305 R/hr Any condition in the opinion of the STED/SED Any condition in the opinion of the STED/SED Any condition in the opinion of the STED/SED Any condition in the opinion of the STED/SED Any condition in the opinion of the STED/SED Any condition in the opinion of the STED/SED

7. STED/SED Judgment that indicates a Potential Loss of the Fuel Clad that indicates a Potential Loss of the RCS that indicates a Loss of the Containment that indicates a Potential Loss of the that indicates a Loss of the Fuel Clad Barrier. that indicates a Loss of the RCS Barrier.

Barrier. Barrier. Barrier. Containment Barrier.

Barrier Status General Emergency Alert Site Area Emergency FG1 - Loss of ANY Two Barriers AND Loss FA1 - ANY Loss or Potential Loss FS1 - Loss or Potential Loss of ANY Two Barriers or Potential Loss of Third Barrier of EITHER Fuel Clad or RCS Barriers Fuel Clad Loss Enter Fuel Clad Potential Loss Enter RCS Loss Enter RCS Potential Loss Enter Containment Loss Enter Containment Potential Loss Enter Emergency Classification GE GE GE GE SAE SAE SAE SAE SAE SAE SAE SAE SAE SAE SAE SAE ALERT ALERT ALERT ALERT NOTE 1: Refer to ER 1.1, Section 1.1, Discussion concerning the proper use of CSFSTs as EALs ER 1.1C Rev. 47