ML16351A319

Revision 16-5 to Emergency Plan

ML16351A319
Person / Time
Site:	McGuire, Mcguire
Issue date:	12/07/2016
From:	Capps S D Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16351A319 (312)

Text

December 7, 2016 Serial No: MNS-16-087 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369, 50-370 Emergency Plan, Revision 16-5 Steven D. Capps Vice President McGuire Nuclear Station Duke Energy MGOlVP I 12700 Hagers Ferry Road Huntersville, NC 28078 0: 980.875.4805 f: 980.875.4809 Steven.Capps@duke-energy.com 10 CFR 50.54(q) Please find attached Revision 16-5 to the McGuire Nuclear Station Emergency Plan. This revision is submitted in accordance with the requirements of 10 CFR 50.54(q) and does not result in a reduction in the effectiveness of the Emergency Plan or the Emergency Plan Implementing Procedures.

Questions regarding this submittal should be directed to George Murphy, McGuire Regulatory Affairs, at (980) 875-5715. Steven D. Capps Attachments www.duke-energy.com I_ U. S. Nuclear Regulatory Commission December 7, 2016 Page 2 (Two Copies) cc: Catherine Haney, Regional Administrator, Region II U. S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 (One Copy) Director Office of Nuclear Material Safety and Safeguards Mail Stop 14 A12 Washington, D. C. 20555-0001 (w/o attachments)

Michael Orenak, Interim Project manager michael.orenak@nrc.gov Mr. Andy Hutto NRC Senior Resident Inspector McGuire Nuclear Station \

DUKE ENERGY McGUIRE NUCLEAR SITE EMERGENCY PLAN 5-/zy,,,lJ. SITE VICE PRESIDENT DATE APPROVED:

____ /1....,....jl_wL

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1to _____ _ REVISION 16-5: November, 2016 EFFECTIVE DATE: November, 2016 ORIGINAL DATE: August 25, 1980 August 25, 1980 Date Issued Change 1, October, 1980 Change 2, February, 1981 Change 3, June, 1981 Change 4, August, 1981 Revision 1, November 16, 1981 Revision 2, February, 1982 Revision 3, February, 1982 Revision 4, April, 1982 Revision 5, June, 1982 Revision 6, July, 1982 Revision 7, September, 1982 Revision 8, November, 1982 Revision 9, January, 1983 Revision 10, February, 1983 Revision 11, June, 1983 Revision 12, November, 1983 Revision 13, March, 1984 Revision 14, August, 1984 Revision 15, January, 1985 Revision 16, March, 1985 Revision 17, May, 1985 Revision 18, November, 1985 Revision 19, January, 1986 Revision 20, July, 1986 Revision 21, May, 1987 Revision 22, June, 1987 Revision 23, November, 1987 Revision 24, March, 1988 Revision 25, July, 1988 Revision 26, July, 1989 Revision 27, September, 1989 Revision 28, October, 1989 Revision 29, November, 1989 Revision 30, March, 1990 Revision 31, April, 1991 Revision 32, July, 1991 Revision 33, September, 1991 Revision 34, October, 1991 Revision 35, December, 1991 Revision 36, January, 1992 DUKE ENERGY COMPANY McGUIRE NUCLEAR SITE EMERGENCY PLAN REVISION LIST 1 Revision 3 7, March 1992 Revision 92-1, August 1992 Revision 92-2, October 1992 Rev. 93-1, April 1993 Rev. 93-2, June, 1993 Rev. 93-3, December 1993 Rev. 94-1, January, 1994 Rev. 94-2, June, 1994 Rev. 94-3, August 1994 Rev. 94-4, October 1994 Rev. 95-1, February 1995 Rev. 95-2, April 1995 Rev. 96-1, April 1996 Rev. 96-2, July 1996 Rev. 97-1, April 1997 Rev. 97-2, May 1997 Rev. 97-3, July, 1997 Rev.98-1,January, 1998 Rev. 98-2, February, 1998 Rev.98-3,May, 1998 Rev. 98-4, July, 1998 Rev. 98-5, August, 1998 Rev. 98-6, November, 1998 Rev. 99-1, March, 1999 Rev. 99-2, July, 1999 Rev. 99-3 November, 1999 Rev. 00-1, April, 2000 Rev. 00-2, May, 2000 Rev. 00-3, November, 2000 Rev. 01-1, January, 2001 Rev. 01-2, June, 2001 Rev. 02-1, March, 2002 Rev. 02-2, August, 2002 Rev. 03-1, April, 2003 Rev. 03-2, June, 2003 Rev. 04-1, February, 2004 Rev. 04-2, July, 2004 Rev. 05-1, July, 2005 Rev. 06-1, January, 2006 Rev. 06-2, September, 2006 Rev. 07-1, May, 2007 Rev. 16-5 November, 2016 Rev. 07-2, December, 2007 Rev. 08-1, September, 2008 Rev. 09-1, July, 2009 Rev. 09-2, December, 2009 Rev. 10-1,l\1ay,2010 Rev. 10-2, November, 2010 Rev. 11-1, l\1arch, 2011 Rev. 11-2, August, 2011 Rev. 11-3, October, 2011 Rev. 12-l,l\1ay,2012 Rev. 12-2, June, 2012 Rev. 12-3, November, 2012 Rev. 12-4, December, 2012 Rev. 13-1, l\1arch, 2013 Rev. 13-2, June, 2013 Rev. 13-3, October, 2013 Rev. 14-1, January, 2014 Rev. 14-2, June 2014 Rev. 14-3, September, 2014 Rev. 14-4, October, 2014 Rev. 14-5, December, 2014 Rev. 15-l,l\1ay,2015 Rev. 15-2, July, 2015 Rev. 15-3, October, 2015 Rev. 16-1,January,2016 Rev. 16-2, April, 2016 Rev. 16-3, September, 2016 Rev. 16-4, November, 2016 Rev. 16-5, November, 2016 2 Rev. 16-5 November, 2016 McGuire Emergency Plan List of Effective Pages Emergency Plan Approval Cover Sheet Coversheet Rev. 16-5 Emergency Plan Revision List Page 1 thru 2 Rev. 16-5 Table of Contents Page 1thru6 List of Figures Page 1thru2 Introduction Pages i-1 thru i-7 Rev. 16-5 Rev. 16-4 Rev. 13-3 A. Assignment of Responsibility Pages A-1 thru A-5 Rev. 16-2 B. Onsite Emergency Organization Pages B-1 thru B-13 Rev. 16-3 C. Emergency Response Support & Resources Pages C-1 thru C-2 Rev. 09-1

• D. Emergency Class System/EAL Basis Document Pages D-1 thru D-259 Rev. 16-5 E. Notification Methods & Procedures Pages E-1 thru E-11 Rev. 16-3 November, 2016 November, 2016 November, 2016 November, 2016 October, 2013 April, 2016 September, 2016 July, 2009 November, 2016 September, 2016 Rev. 16-5 November, 2016 McGuire Emergency Plan List of Effective Pages F. Emergency Communications Pages F-1 thru F-7 Rev. 16-2 G. Public Education

& Information Pages G-1 thru G-4 Rev. 16-2 H. Emergency Facility & Equipment Pages H-1 thru H-17 I. Accident Assessment Pages 1-1 thru 1-3 J. Protective Response Pages J-1 thru J-22 Rev. 16-3 Rev. 16-2

• Rev. 16-4 K. Radiological Exposure Control Pages K-1 thru K-4 Rev. 16-2 L. Medical & Public Health Support Pages L-1 thru L-2 Rev. 13-3 M. Recovery & Re-entry Planning Pages M-1 thru M-5 Rev. 06-2 N. Exercises

& Drills Pages N-1 thru N-3 Rev. 15-1 April, 2016 April, 2016 September, 2016 April, 2016 November, 2016 April, 2016 October, 2013 September, 2006 May,2015 Rev. 16-5 November, 2016 McGuire Emergency Plan List of Effective Pages 0. Radiological Emergency Response Training Pages 0-1thru0-2 Rev. 10-2 P. Development Periodic Review & Distribution of Emergency Plans Pages P-1 thru P-10 Q. Appendices Index Pages Q-1 Appendix 1 Definitions Pages 1 thru 4 Rev. 15-1 Rev. 15-2 Rev. 15-2 Appendix 2 Meteorological Program Pages 1 thru 4 Rev. 15-3 Appendix 3 Alert & Notifications System Rev. 15-2 Pages 1 thru 5 Appendix 4 Evacuation Time Estimates Rev. 15-2 Page 1 Appendix 5 Agreement Letters Rev. 15-3 Pages 1 thru 4 Appendix 6 Emergency Plan Distribution Pages 1 thru 4 Rev. 15-2 November, 2010 May,2015 July, 2015 July, 2015 October, 2015 July, 2015 July, 2015 October, 2015 July, 2015 Rev. 16-5 November, 2016 Appendix 7 SPCC Plan Coversheet Table of Contents Pages 1 thru 75 McGuire Emergency Plan List of Effective Pages Rev. 15-1 Rev. 15-1 Rev. 15-1 Appendix 8 Hazardous Waste Contingency Plan Pages 1 thru 19 Rev. 13-1 Appendix 9 Hazardous Materials Response Plan Pages 1 thru 14 Rev. 13-3 May,2015 May,2015 May,2015 March, 2013 October, 2013 Rev. 16-5 November, 2016 DUKE ENERGY MCGUIRE NUCLEAR SITE EMERGENCY PLAN Table of Contents i. Introduction A. Purpose B. Scope C. Planning Basis A. Assignment of Responsibility Al.a Organization A.Lb Concept of Operations A.1.c Block Diagram Interrelationships A.1.d Key Decisionmaking A.1.e 24 Hour Emergency Response A.2.a Responsibility for and Functions of Emergency Response Organization A.2.b Legal Basis for Authority A.3 Agreement Letters for Emergency Response Support A.4 Individuals Responsible for Continuity of Resources B. On-site Emergency Organization B.1 Plant Staff Under Emergency Conditions B.2 Emergency Coordinator B.3 Emergency Coordinator (line of succession)

B.4 Functional Responsibilities of Emergency Coordinator B.5 Minimum Staffing Requirements B.6 On-site Functional Area Interfaces B.7 Augmented Support of on.:site Emergency Organization B.8 Contractor and Private Organizations B.9 Local Agency Support Services B.9.a Law Enforcement, Emergency Traffic Control, Related Police Matters B.9.b Early W aming or Evacuation of the Populace B.9.c Radiological Emergency Monitoring Assistance B.9.d Hospitals, Medical Support B.9.e Ambulance Service B.9.f Fire-fighting B.9.g Public Health and Safety, Evaluation of the Radiological Situation B.9.h Local, State and Federal Support Responsibilities 1 Page# i-1 i-1 i-3 A-1 A-3 A-3 A-3 A-4 A-4 A-4 A-4 A-5 B-1 B-1 B-1 B-2 B-2 B-2 B-2 B-3 thru B-4 B-4 B-4 B-5 B-5 B-5 B-5 B-5 B-5 B-6 Rev. 16-5 November, 2016 Table of Contents C. Emergency Response Support and Resources C.l.a C.l.b C.1.c C.2.a C.2.b C.3 C.4 Individuals Authorized to Request Federal Assistance Federal Resources Arrival Time Emergency Operations Facility Resources Available to Federal Response Organizations State and County Representation at the Emergency Operations Facility (EOF) Licensee Representation at the Off-Site EOC's Radiological Laboratories-Availability and Capability Emergency Support From Other Organizations D. Emergency Classification System/EAL Basis Document Emergency Classification System/EAL Basis Document E. Notification Methodology F. E.1 E.2 E.2.a E.2.b E.2.c E.2.d E.3 E.4 E.5 E.6 E.7 Notification of Response Organization Activation of Emergency Organization Notification of Unusual Event Alert Site Area Emergency General Emergency Emergency Message Format (Initial)

Emergency Message Format (Follow-up)

State and Local Organizations-Disseminating Public Information Alert and Notification System Supporting Information for Public Information Message Emergency Communications F.l.a 24 Hour Notification Capability F.l.b Communications With State/Local Governments F.l.c Communications With Federal Organizations F.l.d Communications Between Site, EOF, EOC's and Monitoring Team F.l.e Activation of Emergency Personnel F.l.f Communications Between NRC, EOC and Monitoring Teams F.l.g ERDS Data Transfer F.2 Medical Support Communications F.3 Communications System Testing Page# C-1 C-1 C-1 C-1 C-1 C-2 C-2 D-1 thru D-259 E-1 E-1 E-1 E-2 E-4 E-6 E-8 E-9 E-9 E:...9 E-9 F-1 F-1 F-2 F-2 F-2 F-2 F-2 F-3 F-3 Rev. 16-5 November, 2016 Table of Contents G. Public Education and Information G.l/G.2 G.3.a G.3.b G.4.a G.4.b G.4.c G.5 Public Education and Information Program News Group -Location and Contacts News Group -Media Center Public Spokesperson Spokesperson Information Exchange Rumor Control News Media Training Sessions H. Emergency Facilities and Equipment H.1 H.l.a H.l.b H.l.c H.1.d H.2 H.3 H.4 H.5 H.5.a H.5.b H.5.c H.5.d H.6 H.7 H.8 H.9 H.10 H.11 H.12 Technical Support Center/Operations Support Center Control Room Technical Support Center (TSC) Operations Support Center (OSC) Alternate Facilities Emergency Operations Facility (EOF) State and Local Government Emergency Operations Center Activation and Staffing Assessment Actions Meteorological, Hydrologic and Seismic Radiological Monitors Plant Parameters Fire Detection Data, Monitoring Equipment and Analysis Facilities Off-site Radiological Monitoring Meteorology Instrumentation and Procedures Operations Support Center Emergency Equipment/Instrumentation Inspection, Inventory, Operational Check, Calibration Emergency Kits Receipt and Analysis of Field Monitoring Data I. Accident Assessment 1.1 1.2 Emergency Action Level Procedures On-site Capability and Resources to Provide Initial Values and Continuing Assessment Post Accident Sampling Radiation and Effluent Monitors In-plant Iodine Instrumentation Method for Determining Release Source Term Effluent Monitor Readings Vs On-site/Off-site Exposure Meteorological Information Availability Page# G-1 G-1 G-1 G-2 G-2 G-2 G-2 H-1 H-1 H-1 H-2 H-2 .. H-2 thru H-3 H-4 H-4 H-4 H-4 thru H-5 H-6 H-6 H-6 H-6 H-7 H-7 H-7 H-7 H-7 H-7 I-1 I-1 I-1 I-1 I-2 I-2 1.2.a 1.2.b 1.2.c 1.3.a/ 1.3.b I.4 1.5 I.6 Release Rates/Projected Doses for Offscale Instrumentation I-2 I-2 I-2 3 Rev. 16-5 November, 2016 Table of Contents I.7/ Field Monitoring Within EPZ I.8 I.9 Detect and Measure Radioiodine Concentration in the EPZ I.10 Relationship Between Contamination Levels and Integrated Dose/Dose Rates I.11 Plume Tracking J. Protective Response J.l.a Onsite Alerting and Notification thru J.l.d J.2 Evacuation Routes and Transportation J.3 Personnel Monitoring J.4 Site Evacuation Procedures

-Decontamination J.5 Personnel Accountability J.6 Protective Equipment Breathing Apparatus, Protective Clothing, KI J.7 Protective Action Recommendations J.8 Evacuation Time Estimates J.9 Implementing Protective Measures J.10 Implementation of Protective Measures for Plume Exposure Pathway J.10.a EPZMaps J.10.b EPZ Population Distribution Map J.10.c EPZ Population Alerting and Notification J.10.d EPZ Protecting Immobile Persons J.10.e Use ofRadioprotective Drugs for Persons in EPZ J.10.f Conditions for Use of Radioprotective Drugs J.10.g State/County Relocation Plans J.10.h Relocation Center Locations J.10.i Evacuation Route -Traffic Capacities J.10.j Evacuated Area Access Control J.10.k Planning for Contingencies in Evacuation J.10.l State/County Evacuation Time Estimates J.10.m Bases for Protective Action Recommendations J.11 Ingestion Pathway Planning J.12 Relocation Center -Registering and Monitoring 4 Page# I-3 I-3 I-3 I-3 J-1 J-1 J-1 J-2 J-2 J-2 J-3 J-4 J-4 J-5 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-6 J-7 J-7 Rev. 16-5 November, 2016 Table of Contents K. Radiological Exposure Control K.l K.2 K.3 K.3.a K.3.b K.4 K.5 Onsite Exposure Guidelines Doses in Excess of 1 OCFR Part 20 Emergency Personnel Exposure and Records Distribution of Dosimetry Dose Records State/Local Plan for Authorizing Doses Exceeding PA G's Decontamination Page# K.5.a K.5.b K.6 K.6.a

• K.6.b K.6.c K.7 Action Levels for Determining the Need for Decontamination Radiological Decontamination K-1 K-1 K-1 K-1 K-2 K-2 K-2 K-2 K-2 K-2 K-2 K-2 K-3 K-3 Contamination Control Measures Area Access Control Drinking Water and Food Supplies Recovery Efforts Decontamination of Personnel at Relocation Assembly Area L. Medical and Public Health Support L.1 L.2 L.3 L.4 Hospital and Medical Support On-site First Aid Capability Public, Private, Military Hospitals, Emergency Medical Facilities Transport of Accident Victims M. Recovery and Reentry Planning and Post-Accident Operations M.1 M.1.a M.1.b M.2 M.3 M.4 Recovery/Reentry Plans and Procedures Outline of Site Recovery Plans Outline of Recovery Plans Recovery Organization Information to Members of Recovery Organization Total Population Exposure Estimates N. Exercises and Drills N.1.a N.1.b N.2 N.2.a N.2.b N.2.c N.2.d N.2.e N.3 N.4 N.5 Exercises Exercise Scenario/Response Drills Communications Fire Drills Medical Emergency Drills Radiological Monitoring Drills Radiation Protection Drills Exercise and Drill Execution Exercise Critique Critique Action Items 5 L-1 L-1 L-1 L-2 M-1 M-1 M-2 M-3 M-4 M-4 N-1 N-1 N-1 N-1 N-2 N-2 N-2 N-2 N-3 N-3 N-3 Rev. 16-5 November, 2016 Table of Contents 0. Radiological Emergency Response Training 0.1 Offsite Agency Training 0.1.a Emergency Response Training (Offsite Agency) 0.1.b Off-site Support Agency -Participation in Training 0.2 Site Organization Training 0.3 First Aid Training 0.4 Training For Radiological Emergency Response Personnel 0.5 Training Period P. Responsibility for the Planning Effort P.1 Emergency Planning Staff Training P.2 Emergency Response Planning P.3 Site Emergency Planning Manager P.4 Review of Emergency Plan P.5 Distribution of Revised Plans P.6 Supporting Plans P.7 Implementing Procedures P.8 Table of Contents P.9 Audit of Emergency Plan P.10 Telephone Number Updates Q. Appendices Index Appendix 1 Definitions Appendix 2 Meteorological Program Appendix 3 Alert and Notification System Description Appendix 4 Evacuation Time Estimates Appendix 5 Agreement Letters Appendix 6 McGuire Nuclear Site Emergency Plan Distribution Appendix 7 SPCC Plan Appendix 8 Hazardous Waste Contingency Plan Appendix 9 Hazardous Materials Response Plan 6 Page# 0-1 0-1 0-1 0-1 0-1 0-2 0-2 P-1 P-1 P-1 P-1 P-1 P-2 P-2 P-2 P-2 P-3 Rev. 16-5 November, 2016

MCGUIRE NUCLEAR STATION EMERGENCY ACTION LEVEL TECHNICAL BASES D-1 Revision 16-5 November, 2016 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ......................................................................................................................................

3 2.0 DISCUSSION

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3 2.1 Background

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3 2.2 Fission Product Barriers ..........................................................................................................

4 2.3 Fission Product Barrier Classification Criteria ..........................................................................

4 2.4 EAL Organization

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5 2.5 Technical Bases Information

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7 2.6 Operating Mode Applicability

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8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS

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9 3.1 General Considerations

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9 3.2 Classification Methodology

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10

4.0 REFERENCES

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14 4.1 Developmental

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• ... 14 4.2 Implementing

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14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS

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15 6.0 MNS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE

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23 7.0 ATTACHMENTS

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27 1 Emergency Action Level Technical Bases ...................................................................

28 Category R Abnormal Rad Release I Rad Effluent .............................................

28 Category C Cold Shutdown I Refueling System Malfunction

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71 Category H Hazards ...........................................................................................

108 Category S System Malfunction

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153 Category E ISFSI ...............................................................................................

194 Category F Fission Product Barrier Degradation

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198 2 Fission Product Barrier Loss I Potential Loss Matrix and Bases ........................................................................................................

200 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases .......................................

254 D-2 Revision 16-5 November, 2016 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for McGuire Nuclear Station (MNS). It should be used to facilitate review of the MNS EALs and provide historical documentation for future reference.

Decision-makers responsible for implementation of RP/O/A/5700/000 Classification of Emergency, may use this document as a technical reference in support of EAL interpretation.

This information may assist the Emergency Coordinator/EDF Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification making (e.g., the Emergency Coordinator refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).

2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the MNS Emergency Plan. In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007)

Revisions 4 and 5 were subsequently issued for industry implementation.

Enhancements over earlier revisions included:

• Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

• Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).

• Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML 12326A805) (ref.4.1.1

), MNS conducted an EAL implementation upgrade project that produced the EALs discussed herein. D-3 Revision 16-5 November, 2016 2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.

This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss"

• threshold means the barrier no longer assures containment of radioactive materials.

A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier. The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier is the zircalloy tubes that contain the fuel pellets. B. Reactor Coolant System (NCS): The NCS Barrier includes the NCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment (CMT): 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 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Any loss or any potential loss of either Fuel Clad or NCS barrier Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss of the third barrier D-4 Revision 16-5 November, 2016 2.4 EAL Organization

'The MNS EAL scheme includes the following features:

• Division of the EAL set into three broad groups: o EALs applicable under fill.plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered.

o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode. o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or No Mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition.

This approach significantly minimizes the total number of EALs that must.be reviewed by the user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

• Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.

The MNS EAL categories are aligned to and represent the NEI 99-01"Recognition Categories." Subcategories are used in the MNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds.

The MNS EAL categories and subcategories are listed below. D-5 Revision 16-5 November, 2016 EAL Groups, Categories and Subcategories EAL Group/Category R -Abnormal Rad Levels I Rad Effluent H -Hazards and Other Conditions Affecting Plant Safety E -Independent Spent Fuel Storage Installation (ISFSI) S -System Malfunction F -Fission Product Barrier Degradation C -Cold Shutdown I Refueling System Malfunction EAL Subcategory 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels 1 -Security 2 -Seismic Event 3 -Natural or Technological Hazard 4-Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -Emergency Coordinator Judgment 1 -Confinement Boundary 1 -Loss of Essential AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4-NCS Activity 5 -NCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Containment Failure 9 -Hazardous Event Affecting Safety Systems None 1 -NCS Level 2 -Loss of Essential AC Power 3 -NCS Temperature 4 -Loss of Vital DC Power 5 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration.

The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.

D-6 Revision 16-5 November, 2016 2.5 Techr:iical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory.

A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category.

For each EAL, the following information is provided:

Categorv Letter & Title Subcategorv Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle)

Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel.

Four characters define each EAL identifier:

1. First character (letter):

Corresponds to the EAL category as described above (R, C, H, S, For E) 2. Second character (letter):

The emergency classification (G, S, A or U) G = General Emergency S = Site Area Emergency A= Alert U =Unusual Event 3. Third character (number):

Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1 ). If a category does not have a subcategory, this character is assigned the number one (1 ). 4. Fourth character (number):

The numerical sequence of the EAL within the EAL subcategory.

If the subcategory has only one EAL, it is given the number one (1 ). Classification (enclosed in rectangle):

Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix D-7 Revision 16-5 November, 2016 Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, NM -No Mode, or All. (See Section 2.6 for operating mode definitions)

Definitions:

If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: A basis section that provides MNS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. MNS Basis Reference(s):

Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.7) 1 Power Operation Keff 2:.,.0.99 and reactor thermal power> 5% 2 Startup Keff ::::_0.99 and reactor thermal power 3 Hot Standby Keff < 0.99 and average coolant temperature 2:..350°F 4 Hot Shutdown Keff < 0.99 and average coolant temperature 350°F > T avg > 200 °F 5 Cold Shutdown Keff < 0.99 and average coolant temperature 6 Refueling One or more reactor vessel head closure bolts are less than fully tensioned NM No mode Reactor vessel contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

D-8 Revision 16-5 November, 2016 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerati,ons When making an emergency classification, the Emergency Coordinator/EOF Director 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 based on loss or potential loss of Fission Product Barrier Thresholds.

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.12). 3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions.

A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy.

For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator/EOF Director 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.

D-9 Revision 16-5 November, 2016 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component.

In these cases, the controls associated with the planning, preparation and execution of the work will ensure that complian,ce 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 (ref. 4.1.4). 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, NCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis.

In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available).

The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary.

The NEI 99-01 EAL scheme provides the Emergency Coordinator/EDF Director 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 Emergency Coordinator/EDF Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition.

A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.

The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started. When assessing an EAL that specifies a time duration for the 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 (ref. 4.1.14). D-10 Revision 16-5 November, 2016 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable EGL identified during this review is declared.

For example:

• If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.

There is no "additive" effect from multiple EALs meeting the same EGL. For example:

• If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRG Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2). 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.

If an event or condition occurs, and results in a mode change before the emergency is declared, the

• emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared).

Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.

In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Coordinator/EDF Director 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 EGL is IMMINENT).

If, in the judgment of the Emergency Coordinator/EOF Director, 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.

D-11 Revision 16-5 November, 2016 3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2). 3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.

By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.

If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip. 3.2.6 Classification of Transient Conditions Many of the I Cs and/or EALs employ time-based criteria.

These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.

In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes).

The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response -In instances 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.

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 A TWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and NCS barriers).

If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. D-12 Revision 16-5 November, 2016 Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator/EDP Director 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.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition.

This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event orcondition no longer exists at the time of discovery.

This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable.

Specifically, the event should be reported to the NRG in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition.

The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRG is discussed in NUREG-1022 (ref. 4.1.3). D-13 Revision 16-5 November, 2016

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML 12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines:

10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 MNS UFSAR Figure 2-4 Plot Plan and Site Area 4.1.7 Technical Specifications Table 1.1-1 Modes 4.1.8 PT/1 (2)/A/4200/002 C (Containment Closure) 4.1.9 PRO-NGGC-0201 NGG Procedure Writers Guide 4.1.10 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.11 MNS ISFSI Certificate of Compliance 4.1.12 MNS Emergency Plan 4.1.13 MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary 4.2 Implementing 4.2.1 RP/O/A/5700/000 Classification of Emergency 4.2.2 NEI 99-01 Rev. 6 to MNS EAL Comparison Matrix 4.2.3 MNS EAL Matrix D-14 Revision 16-5 November, 2016 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted) Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.

The definitions of these terms are provided below. Alert Events are in progress, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the MNS ISFSI, Confinement Boundary is defined as the Transportable Storage Cask (TSC) for TN, UMS and MAGNASTOR storage systems. Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to MNS, Containment Closure is established when the requirements of PT/1 (2)/A/4200/002 Care met (ref. 4.1.8). Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRG) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

• Unusual Event (UE)

• Alert

• Site Area Emergency (SAE)

• General Emergency (GE) D-15 Revision 16-5 November, 2016 EPAPAGs Environm'ent Protection Agency Protective Action Guidelines.

The EPA PAGs are expressed in terms of dose commitment:

1 Rem TEDE or 5 Rem COE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires MNS to recommend protective actions for the general public to offsite planning agencies.

Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events require a post-event inspectiqn to determine if the attributes of an explosion are present. Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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

Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile actions that result 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. Hostage. A person(s) held as leverage against the station to ensure that demands will be met by the station. D-16 Revision 16-5 November, 2016 Hostile Action An act toward MNS 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 Other acts that satisfy the overall intent may be inclyded.

Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on MNS. 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. lmpede(d)

Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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.

Intrusion The act of entering without authorization.

Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force. Maintain Take appropriate action to hold the value of an identified parameter within specified limits. Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value. Owner Controlled Area Area outside the PROTECTED AREA fence that immediately surrounds the plant. The site property owned by, or otherwise under the control of, Duke Energy. D-17 Revision 16-5 November, 2016 Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Protected Area An area encompassed by physical barriers and to which access is controlled; The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area (ref. 4.1.6). NCS Intact The NCS should be considered intact when the NCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway. Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits 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 related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

D-18 Revision 16-5 November, 2016 Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundary Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary (ref. 4.1.13). Unisolable An open or breached system line that cannot be isolated, remotely or locally. Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Unusual Event Events are in progress or have occurred which indicate a potential degradation 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. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

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.

D-19 Revision 16-5 November, 2016 5.2 Abbreviations/Acronyms

°F ******************************************;************************************************************Degrees 0 ********************************************

(..............................................................................

DeQrees AC .......................................................................................................

Alternating Current AP ....................................................................................

Abnormal Operating Procedure A TWS ................................... ..................................

Anticipated Transient Without Scram MNS ............................................................................................

McGuire Nuclear Station COE .......................................................................................

Committed Dose Equivalent CFR .....................................................................................

Code of Federal Regulations CSFST .......................................................................

Critical Safety Function Status Tree OBA ...............................................................................................

Design Basis Accident DC ...............................................................................................................

DirectCurrent EAL .............................................................................................

Emergency Action Level EC ...............................................................................................

Emergency Coordinator EGGS ............................................................................

Emergency Core Cooling System EGL. .................................................................................

Emergency Classification Level EOF ..................................................................................

Emergency Operations Facility EOP ...............................................................................

Emergency Operating Procedure EPA ..............................................................................

Environmental Protection Agency ERG ................................................................................

Emergency Response Guideline EPIP ................................................................

Emergency Plan Implementing Procedure ESF .........................................................................................

Engineered Safety Feature FAA ..................................................................................

Federal Aviation Administration FBI ...................................................................................

Federal Bureau of Investigation FEMA. ..............................................................

Federal Emergency Management Agency FSAR ....................................................................................

Final Safety Analysis Report GE .....................................................................................................

General Emergency IC ..........................................................................................................

Initiating Condition IPEEE .................

Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI. ...........................................................

Independent Spent Fuel Storage Installation Ketr .........................................................................

Effective Neutron Multiplication Factor LCO ..................................................................................

Limiting Condition of Operation LER ................................................................................................

Licensee Event Report LOCA .........................................................................................

Loss of Coolant Accident D-20 Revision 16-5 November, 2016 LWR ...................................................................................................

Light Water Reactor MPG...................................

Maximum Permissible Concentration/Multi-Purpose Canister MSIV .......................................................................................

Main Steam Isolation Valve ' MSL ........................................................................................................

Main Steam Line mR, mRem, mrem, mREM ..............................................

milli-Roentgen Equivalent Man MW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Megawatt NCS ............................................................................................

Reactor Coolant System NEI ..............................................................................................

Nuclear Energy Institute NESP ...................................................................

National Environmental Studies Project NPP ..................................................................................................

Nuclear Power Plant NRC ................................................................................

Nuclear Regulatory Commission NSSS ................................................................................

Nuclear Steam Supply System NORAD ...................................................

North American Aerospace Defense Command (NO)UE ................................................................................

Notification of Unusual Event OBE ......................................................................................

Operating Basis Earthquake OCA ...............................................................................................

OwnerControlled Area ODCM ............................................................................

Off-site Dose Calculation Manual ORO .................................................................................

Offsite Response Organization PA..............................................................................................................

Protected Area PAG ........................................................................................

Protective Action Guideline PRA/PSA .....................

Probabilistic Risk Assessment I Probabilistic Safety Assessment PWR .......................................................................................

Pressurized Water Reactor PSIG ................................................................................

Pounds per Square Inch Gauge R ........................................................................................................................

Roentgen Rem, rem, REM .......................................................................

Roentgen Equivalent Man RETS .........................................................

Radiological Effluent Technical Specifications RPS ........................................................................................

Reactor Protection System RV .............................................................................................................

Reactor Vessel RVLIS .................................................................

Reactor Vessel Level Indicating System SAR ...............................................................................................

Safety Analysis Report SBGTS .........................................................................

Stand-By Gas Treatment System SBO . ... .. . .. . . .. .. . .. . . .... .... .. .. ... . ... ... ... . ... .. .. . . ... . ...... .. ... ....... .... ..... ... .. . .. . ..... .. .. . Station Blackout SCBA. .......................

'...... .. .. ... ... . . . . .. .. .. . .. .. . .. .. .. .. . . .. ... Self-Contained Breathing Apparatus SG .........................................................................................................

Steam Generator D-21 Revision 16-5 November, 2016 SI ..............................................................................................................

Safety Injection SLC .................................................................................

Selected Licensee Commitment SPDS ...........................................................................

Safety Parameter Display System SRO ............................................................................................

Senior Reactor Operator SSF ..........................................................................................

StandbyShutdown Facility TEDE ...............................................................................

Total Effective Dose Equivalent TOAF ....................................................................................................

Top of Active Fuel TSC ..........................................................................................

Technical Support Center WOG ..................................................................................

Westinghouse Owners Group D-22 Revision 16-5 November, 2016 6.0 MNS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a MNS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the MNS EALs based on the NEI guidance can be found in the EAL Comparison Matrix. MNS EAL RU1.1 RU1.2 RU2.1 RA1.1 RA1.2 RA1.3 RA1.4 RA2.1 RA2.2 RA2.3 RA3.1 RA3.2 RS1.1 RS1.2 RS1.3 RS2.1 RG1.1 RG1.2 RG1.3 RG2.1 NEI 99-01 Rev. 6 IC Example EAL AU1 1, 2 AU1 3 AU2 1 M1 1 M1 2 M1 3 M1 4 M2 1 M2 2 M2 3 M3 1 M3 2 AS1 1 AS1 2 AS1 3 AS2 1 AG1 1 AG1 2 AG1 3 AG2 1 D-23 Revision 16-5 November, 2016 MNS EAL CU1.1 CU1.2 CU2.1 CU3.1 CU3.2 CU4.1 CU5.1 CA1.1 CA1.2 CA2.1 CA3.1 CA6.1 CS1.1 CG1.1 FA1.1 FS1.1 FG1.1 HU1.1 HU1.2 HU1.3 HU2.1 HU3.1 HU3.2 HU3.3 HU3.4 NEI 99-01 Rev. 6 IC Example EAL CU1 1 CU1 2 CU2 1 CU3 1 CU3 2 CU4 1 CU5 1,2,3 CA1 1 CA1 2 CA2 1 CA3 1, 2 CA6 1 CS1 3 CG1 2 FA1 1 FS1 1 FG1 1 HU1 1 HU1 2 HU1 3 HU2 1 HU3 1 HU3 2 HU3 3 HU3 4 D-24 Revision 16-5 November, 2016 MNS EAL HU4.1 HU4.2 HU4.3 HU4.4 HU7.1 HA1.1 HA1.2 HAS.1 HA6.1 HA7.1 HS1.1 HS6.1 HS7.1 HG1.1 HG7.1 SU1.1 SU3.1 SU4.1 SU4.2 SUS.1 SU6.1 SU6.2 SU7.1 SU8.1 SA1.1 NEI 99-01 Rev. 6 IC Example EAL HU4 1 HU4 2 HU4 3 HU4 4 HU7 1 HA1 1 HA1 2 HAS 1 HA6 1 HA? 1 HS1 1 HS6 1 HS7 1 HG1 1 HG7 1 SU1 1 SU2 1 SU3 1 SU3 2 SU4 1, 2, 3 SUS 1 SUS 2 SU6 1, 2, 3 SU7 1, 2 SA1 1 D-25 Revision 16-5 November, 2016 MNS EAL SA3.1 SA6.1 SA9.1 SS1.1 SS2.1 SS6.1 SG1.1 SG1.2 EU1.1 NEI 99-01 Rev. 6 IC Example EAL SA2 1 SAS 1 SA9 1 SS1 1 SS8 1 SSS 1 SG1 1 SG8 1 E-HU1 1 D-26 Revision 16-5 November, 2016 7.0 ATTACHMENTS 7 .1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis D-27 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category R -Abnormal Rad Release I Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms.

Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases.

At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety. Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits. 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

D-28 Revision 16-5 November, 2016 Category:

ATTACHMENT 1 EAL Bases Subcategory: R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the SLC limits for 60 minutes or longer EAL: RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor> column "UE" for ;::: 60 min. (Notes 1, 2, 3) Note 1: Note 2: Note 3: Table R-1 Release Point The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit 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 the specified time limit. If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Effluent Monitor Classification Thresholds Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1(2)EMF36L

---

4.85E+6 cpm 13.1 OE+3 cpm Ill ::::s 0 Cl) Unit Vent Noble Gas High Ill ca Cl Liquid Waste Effluent Line High "t:I *s CVUCDT High er :J Mode Applicability:

All Definition{s):

None Basis: 1(2)EMF36H EMF49H 1(2)EMF44H 12.61 E+4 cpm 2.61E+3 cpm 2.70E+2 cpm ------------2.15E+2 cpm -----------4.29E+2 cpm The column "UE" gaseous and liquid release values in Table R-1 represent two times the appropriate SLC release rate limits associated with the specified monitors (ref. 2, 3, 4). Gaseous Releases Instrumentation that may be used to assess this EAL is listed below (ref. 1 ):

• Unit Vent Noble Gas Low Monitor-1 (2)EMF36L D-29 Revision 16-5 November, 2016 Liquid Releases ATTACHMENT 1 EAL Bases Instrumentation that may be used to assess this EAL is listed below (ref. 1 ):

• Liquid Waste Effluent Line High Monitor-EMF49H (batch release)

• CVUCDT High Monitor -1 (2)EMF44H This IC addresses a potential decrease in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).

It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.

Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.

The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

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

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

Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.

Escalation of the emergency classification level would be via IC RA 1. D-30 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. MNS ODCM Section 3.0 Setpoint Calculations

2. MNS-SLC 16.11.1 Liquid Effluents

-Concentration

3. MNS-SLC 16.11.6 Dose Rate -Gaseous Effluents

4. EP-EALCALC-MNS-1401 MNS Radiological Effluent EAL Values, Rev. 0 5. NEI 99-01 AU1 D-31 Revision 16-5 November, 2016 Category:

---

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the SLC limits for 60 minutes or longer. EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x SLC limits 60 min. (Notes 1, 2) Note 1: Note 2: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit 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 the specified time limit. Mode Applicability:

All Definition(s):

None Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).

It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.

Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.

The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. D-32 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc} Escalation of the emergency classification level would be via IC RA 1. MNS Basis Reference(s):

1. MNS Offsite Dose Calculation Manual 2. MNS-SLC 16.11.1 Liquid Effluents

-Concentration

3. MNS-SLC 16.11.6 Dose Rate -Gaseous Effluents

4. NEI 99-01 AU1 D-33 Revision 16-5 November, 2016 Category:

ATTACHMENT 1 EAL Bases Subcategory: R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condit,ion:

Release of gaseous or liquid radioactivity rE;:lsulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: RA1.1 Alert Reading on any Table R-1 effluent radiation monitor> column "ALERT" for 15 min. (Notes 1, 2, 3, 4) Note 1: Note 2: Note 3: Note4 !Table R-1 Release Point The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit 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 the specified time limit. If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Effluent Monitor Classification Thresholds Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1(2)EMF36L

---

14.85E+6 cpm 3.10E+3 cpm en ::::i 0 GI Unit Vent Noble Gas High 1(2)EMF36H

a (!) Liquid Waste Effluent Line High EMF49H "'C *= CVUCDTHigh 1(2)EMF44H er :J Mode Applicability:

All Definition(s):

None 2.61E+4 cpm 2.61E+3 cpm -----------D-34 f2.70E+2 cpm ----------2.15E+2 cpm 4.29E+2 cpm Revision 16-5 November, 2016 Basis: ATTACHMENT 1 EAL Bases This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 10 mRem TEOE

• 50 mRem COE Thyroid The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEOE or COE Thyroid) (ref. 2). Instrumentation that may be used to assess this EAL is Unit Vent Noble Gas Low Monitor -1 (2)EMF36L and Unit Vent Noble Gas High Monitor -1 (2)EMF36H (ref. 1 ). This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

The TEOE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE. 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. MNS Basis Reference(s):

1. MNS OOCM Section 3.0 Setpoint Calculations

2. EP-EALCALC-MNS-1401 MNS Radiological Effluent EAL Values, Rev. 0 3. NEI 99-01 AA1 D-35 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: RA1.2 Alert

---

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS 1.1 and RG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY -Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: Dose assessments are performed by computer-based methods (ref. 1, 2) This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. 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 D-36 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RS1. MNS Basis Reference{s):

1. HP/O/B/1009/029 Initial Response On-Shift Dose Assessment

2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment

3. NEI 99-01 AA1 D-37 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: RA1.3 Alert ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses> 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY-Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1 ). This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. D-38 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Classification based on effluent monitor readings assumes that a release path to the

• environment is established.

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

• Escalation of the emergency classification level would be via IC RS1. MNS Basis Reference(s):

1. MNS Offsite Dose Calculation Manual 2. NEI 99-01 AA1 D-39 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: RA1.4 Alert ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates > 10 mR/hr expected to continue for;::: 60 min.

• Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY-Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: HP/O/B/1009/023, Environmental Monitoring for Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1 ). This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions D-40 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEOE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE. Escalation of the emergency classification level would be via IC RS1. MNS Basis Reference(s):

1. HP/0/8/1009/023 Environmental Monitoring for Emergency Conditions

2. NEI 99-01 AA1 D-41 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL: RS1.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor> column "SAE" for 15 min. (Notes 1, 2, 3, 4) Note 1: Note 2: Note 3: Note 4: Table R-1 Release Point The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit 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 the specified time limit. If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Effluent Monitor Classification Thresholds Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1(2)EMF36L

---

14.85E+6 cpm 3.10E+3 cpm Ill 0 Cl) Unit Vent Noble Gas High 1(2)EMF36H

Q (!) Liquid Waste Effluent Line High EMF49H ,, *:; CVUCDTHigh 1(2)EMF44H er ::i Mode Applicability:

All Definition(s):

None 2.61E+4 cpm 2.61E+3 cpm ' ----------D-42 l2.70E+2 cpm ---------2.15E+2 cpm 4.29E+2 cpm Revision 16-5 November, 2016 Basis: ATTACHMENT 1 EAL Bases This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 100 mRem TEOE

• 500 mRem COE Thyroid The column "SAE" gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEOE or COE Thyroid) (ref. 1). Instrumentation that may be used to assess this EAL is Unit Vent Noble Gas High Monitor -1 (2)EMF36H (ref 2). 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 TEOE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE. 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. MNS Basis Reference(s):

1. EP-EALCALC-MNS-1401 MNS Radiological Effluent EAL Values, Rev. 0 2. MNS OOCM Section 3.0 Setpoint Calculations

3. NEI 99-01 AS1 D-43 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition{s):

SITE BOUNDARY-Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: Dose assessments are performed by computer-based methods (ref. 1, 2) This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greaterthan 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. 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.

D-44 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RG1. MNS Basis Reference(s):

1. HP/O/B/1009/029 Initial Response On-Shift Dose Assessment

2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment

3. NEI 99-01 AS1 D-45 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates > 100 mR/hr expected to continue 60 min.

• Analyses of field survey samples indicate thyroid COE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY -Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: HP/O/B/1009/023, Environmental Monitoring for Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1 ). 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.

D-46 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases The TEOE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE. Escalation of the emergency classification level would be via IC RG1. MNS Basis Reference(s):

1. HP/O/B/1009/023 Environmental Monitoring for Emergency Conditions

2. NEI 99-01 AS1 D-47 Revision 16-5 November, 2016 l Category:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor> column "GE" for 15 min. (Notes 1, 2, 3, 4) Note 1: Note 2: Note 3: Note 4: Table R-1 Release Point The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit 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 the specified time limit. If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Effluent Monitor Classification Thresholds Monitor GE SAE Alert UE Unit Vent Noble Gas Low 1(2)EMF36L

---

14.85E+6 cpm cpm UI ::J 0 Cl) Unit Vent Noble Gas High 1(2)EMF36H Kl (!) Liquid Waste Effluent Line High EMF49H ,, *:; CVUCDTHigh 1(2)EMF44H CT :i Mode Applicability:

All Definition(s):

None 12.61 E+4 cpm 2.61E+3 cpm ------------D-48 2.70E+2 cpm ---------l2.15E+2 cpm 14.29E+2 cpm Revision 16-5 November, 2016 Basis: ATIACHMENT 1 EAL Bases This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 1000 mRem TEOE

• 5000 mRem COE Thyroid The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEOE or COE Thyroid) (ref. 1 ). Instrumentation that may be used to assess this EAL is the Unit Vent Noble Gas High Monitor 1 (2)EMF36H (ref 2). 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 TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE. 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.

MNS Basis Reference(s}:

1. EP-EALCALC-MNS-1401 MNS Radiological Effluent EAL Values, Rev. 0 2. MNS OOCM Section 3.0 Setpoint Calculations

3. NEI 99-01 AG1 D-49 Revision 16-5 November, 2016 _j Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY -Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: Dose assessments are performed by computer-based methods (ref. 1, 2) 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. 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.

D-50 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. HP/O/B/1009/029 Initial Response On-Shift Dose Assessment

2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment

3. NEI 99-01 AG1 D-51 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates> 1,000 mR/hr expected to continue for;:: 60 min.

• Analyses of field survey samples indicate thyroid COE > 5,000 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY -Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary.

Basis: HP/O/B/1009/023, Environmental Monitoring for Emergency Conditions provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1 ). 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.

D-52 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases The TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE. MNS Basis Reference(s):

1. HP/0/8/1009/023 Environmental Monitoring for Emergency Conditions

2. NEI 99-01 AG1 D-53 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Unplanned loss of water level above irradiated fuel RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication AND UNPLANNED rise in corresponding area radiation levels as indicated by EITHER of the following radiation monitors:

• 1EMF17 (2EMF4) Spent Fuel Building Refueling Bridge

• 1EMF16 (2EMF3) Containment Building Refueling Bridge (Mode 6) Mode Applicability:

All Definition(s):

UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY-.

The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway. Basis: The spent fuel pool low water level alarm setpoint is OAC point M1 (2)D2937 (ref. 1 ). Water level restoration instructions are performed in accordance with AOPs (ref. 2, 3). The specified radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 2, 3). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING PATHWAY level are not classifiable under this EAL. 1EMF16 (2EMF3) Containment Building Refueling Bridge monitors are only operable in Mode 6 (Refueling).

When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the reactor vessel and spent fuel pool. D-54 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation.

Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available).

A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

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. MNS Basis Reference(s):

1. OP/1(2)/A/6102/001

2. AP/1 (2)/A/5500/40 Loss of Refueling Cavity Level 3. AP/1 (2)/A/5500/41 Loss of Spent Fuel Cooling or Level 4. NEI 99-01 AU2 D-55 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

All Definition(s):

REFUELING PATHWAY-.

The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway. Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel 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 EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

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

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered.

To the degree possible, readings should be considered in combination with other available indications of inventory loss. 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 RS1 or RS2. D-56 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. AP/1 (2)/A/5500/040 Loss of Refueling Cavity Level 2. AP/1 (2)/A/5500/041 Loss of Spent Fuel Cooling or Level 3. NEI 99-01 AA2 D-57 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND A Trip 2 radiation alarm on any of the following radiation monitor indications:

• 1EMF17 (2EMF4) Spent Fuel Building Refueling Bridge

• 1 EMF16 (2EMF3) Containment Building Refueling Bridge (Mode 6)

• 1 EMF42 (2EMF42) Fuel Building Ventilation

• 1 EMF39 (2EMF39) Containment Gas Mode Applicability:

All Definition(s}:

None Basis: The specified radiation monitors are those expected to see increased area radiation levels as a result of damage to irradiated fuel (ref. 1 ). 1EMF16 (2EMF3) Containment Building Refueling Bridge monitors are only operable in Mode 6 (Refueling).

The Trip 2 alarm setpoints for the radiation monitors are set to be indicative of significant increases in area and/or airborne radiation (ref. 2). 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 EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EU1 .1. D-58 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Escalation of the emergency would be based on either Recognition Category R or C I Cs. This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.

A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency classification level would be via IC RS1 or RS2. MNS Basis Reference(s):

1. AP/1 (2)/A/5500/25 Spent Fuel Damage 2. HP/O/B/1003/008 Determination of Radiation Monitor Setpoints (EMFs) 3. NEI 99-01 AA2 D-59 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel EAL: RA2.3 Alert Spent fuel pool level :5 -15 ft. (756 ft. ele.) (KFP5350 or NVPG6530)

Mode Applicability:

All Definition(s):

None Basis: Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1 ), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). The SFP level instruments consist of a primary channel (1 (2)KFP5350) and back-up channel (1 (2)NVPG6530) each spanning approximately 30 ft. (-25 ft. -+5 ft.) (745 ft. ele. -775 ft. ele.). Level 2 is a SFP level of -15 ft. (756' ft. ele.) or approximately 10 ft. above the top of the SFP racks (ref. 2, 3). 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. Escalation of the emergency would be based on either Recognition Category R or C I Cs. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC RS1. D-60 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. NRC EA-12-051 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. MNS-14-023 Second Six-Month Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

3. Engineering Change Packages #109073 and #10907 4 4. NEI 99-01 AA2 D-61 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Spent fuel pool level at the top of the fuel racks RS2.1 Site Area Emergency Spent fuel pool level :s; -25 ft. (746 ft. ele.) (KFP5350 or NVPG6530)

Mode Applicability:

All Definition(s):

None Basis: Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1 ), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). The SFP level instruments consist of a primary channel (1 (2)KFP5350) and back-up channel (1(2)NVPG6530) each spanning approximately 30 ft. (-25 ft. -+5 ft.) (745 ft. ele. -775 ft. ele.). Level 3 is a SFP level of -25 ft. (746' ft. ele.) or approximately the top of the SFP racks (ref. 2, 3). This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

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

Escalation of the emergency classification level would be via IC RG1 or RG2. D-62 Revision 16-5 November, 2016

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ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

1. NRC EA-12-051 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. MNS-14-023 Second Six-Month Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

3. Engineering Change Packages #109073 and #10907 4 4. NEI 99-01 AS2 D-63 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 2 -Irradiated Fuel Event Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer RG2.1 General Emergency Spent fuel pool level cannot be restored to> -25 ft. (746 ft. ele.) (KFP5350 or NVPG6530) 60 min. (Note 1) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

All Definition(s):

None Basis: Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1 ), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). The SFP level instruments consist of a primary channel (1 (2)KFP5350) and back-up channel (1 (2)NVPG6530) each spanning approximately 30 ft. (-25 ft. -+5 ft.) (745 ft. ele. -775 ft. ele.). Level 3 is a SFP level of -25 ft. (746' ft. ele.) or approximately the top of the SFP racks (ref. 2, 3). This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

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

D-64 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

1. NRC EA-12-051 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. MNS-14-023 Second Six-Month Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)

3. Engineering Change Packages #109073 and #10907 4 4. NEI 99-01 AG2 D-65 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 3 -Area Radiation Levels Initiating Condition:

Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.1 Alert Dose rates> 15 mR/hr in EITHER of the following areas: Control Room (1 EMF12) OR Central Alarm Station (by survey) Mode Applicability:

All Definition(s):

IMPEDE(D)

-Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis: Areas that meet this threshold include the Control Room and the Central Alarm Station (CAS). 1EMF Channel 12 monitors the Control room for area radiation (ref. 1). The CAS is included in this EAL because of its' importance to permitting access to areas required to assure safe plant operations.

There is no permanently installed CAS area radiation monitors that may be used to assess this EAL threshold.

Therefore this threshold must be assessed via local radiation survey for the CAS. This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.

As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.

An emergency declaration is not warranted if the following condition applies. D-66 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases

• The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.). Escalation of the emergency classification level would be via Recognition Category R, C or F I Cs. , ' MNS Basis Reference(s):

1. UFSAR Table 12-11 Area Radiation Monitors 2. NEI 99-01 AA3 D-67 Revision 16-5 November, 2016

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ATTACHMENT 1 EAL Bases Category: R -Abnormal Rad Levels I Rad Effluent Subcategory: 3 -Area Radiation Levels Initiating Condition:

1 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table R-2 Safe Operation

& Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area Unit 2 Room/Area Modes Auxiliary 716' P/C, RHole, near 1Nl-185, ABPC thru CAD Door, FF59 4 Outside CAD 212 Auxiliary 750' 800 (1EMXA) 820 {2EMXA) 3, 4 803 (1ETA) 805 (2ETA) 3,4 702 {Elec. Pene.) 713 {Elec. Pene.) 3 Auxiliary 733' 722 (1 EMXB-1) 724 {2EMXB-1) 3,4 705 (1ETB) 716 (2ETB) 3,4 Mode Applicability:

All Definition(s):

JMPEDE(D)

-Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. D-68 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown.

Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included.

In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1 ). This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.

As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.

For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply:

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

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

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

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

• If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. Escalation of the emergency classification level would be via Recognition Category R, C or F I Cs. D-69 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. NEI 99-01 AA3 D-70 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category C -Cold Shutdown I Refueling System Malfunction EAL Group: Cold Conditions (NCS temperature s 200°F); EALs in this category are applicable only in one or more cold operating modes. Category C EALs are directly associated with cold shutdown or refueling system safety functions.

Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown.

Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable.

The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to NCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 -Cold Shutdown, 6 -Refueling, NM -No Mode). The events of this category pertain to the following subcategories:

1. 1. NCS Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite power sources for 4160 VAC essential buses. 3. NCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of power to or degraded voltage on the 125 VDC vital buses. D-71 Revision 16-5 November, 2016

5. Loss of Communications ATTACHMENT 1 EAL Bases Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

D-72 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 1 -NCS Level Initiating Condition:

UNPLANNED loss of NCS inventory for 15 minutes or longer EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in NCS water level less than a required lower limit 15 min. (Note 1) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: NCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.

With the plant in Cold Shutdown, NCS water level is normally maintained above the pressurizer low level setpoint of 17% (ref. 1 ). However, if NCS level is being controlled below the pressurizer low level setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the NCS that is the concern. With the plant in Refueling mode, NCS water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.7 requires at least 23 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2). This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor NCS 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 NCS water inventory are carefully planned and controlled.

An UNPLANNED event that results in water level decreasing below a procedurally required D-73 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This .EAL recognizes that the minimum required NCS 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. Continued loss of NCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Tree -Inventory

2. MNS Technical Specifications Section 3.9.7 Refueling Cavity Water Level 3. NEI 99-01 CU1 D-74 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 1 -NCS Level Initiating Cpndition:

UNPLANNED loss of NCS inventory for 15 minutes or longer EAL: CU1.2 Unusual Event NCS water level cannot be monitored AND EITHER

• UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory

• Visual observation of unisolable NCS leakage fable C-6

• NCDT

• PRT

• CFAEsump

• ND/NSsump

• RHT

• WOT

• WEFT

• SRST Mode Applicability: 5 -Cold Shutdown, 6-Refueling Definition(s):

Sumps/Tanks UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the NCS will normally be intact and standard NCS level monitoring means are available.

NCS level in the Refueling mode is normally monitored using the sight glass. In this EAL, all water level indication is unavailable and the NCS inventory loss must be detected by indirect leakage indications.

Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of NCS leakage. If the make-up rate to the NCS unexplainably rises above D-75 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the NCS that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 1, 2),. This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV 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 NCS water inventory are carefully planned and controlled.

An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL addresses a condition where all means to determine RPV 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 NCS. Continued loss of NCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. MNS Basis Reference(s):

1. AP/1(2)/A/5500/10 NC System Leakage Within the Capacity of Both NV Pumps 2. PT/1 (2)/A/4150/001 D Identifying NC System Leakage 3. NEI 99-01 CU1 D-76 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: CA1.1 Alert ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 1 -NCS Level Loss of NCS inventory Loss of NCS inventory as indicated by NCS water level < 5 in. above hotleg centerline Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

None Basis: 5.1 in. above hotleg centerline (reounded to 5 in.) NCS level indication is the lowest level to assure adequate net positive suction head and prevent ND pump cavitation and air entrainment for all flow rates (ref. 1 ). This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).

This condition represents a potential substantial reduction in the level of plant safety. For this EAL, a lowering of NCS water level below 5 in. above hotleg centerline indicates that operator actions have not been successful in restoring and maintaining NCS 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, this EAL is concerned with the loss of NCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in NCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. If NCS water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. MNS Basis Reference(s):

1. EP Calculation File MCC-1552.08-00-0208

2. NEI 99-01 CA1 D-77 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: CA1.2 Alert ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 1 -NCS Level Loss ofi NCS inventory NCS water level cannot be monitored 15 min. (Note

1) AND EITHER

• UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory

• Visual observation of unisolable NCS leakage Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-6 Sumps/Tanks

• NCDT

• PRT

• CFAEsump

• ND/NS sump

• RHT

• WDT

• WEFT

• SRST Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the NCS will normally be intact and standard RPV level monitoring means are available.

In the Refuel mode, the NCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually.

D-78 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases In this EAL, all NCS water level indication would be unavailable for greater than 15 minutes, and the NCS inventory loss must be detected by indirect leakage indications.

Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the NCS that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 1, 2). This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).

This condition represents a potential substantial reduction in the level of plant safety. For this EAL, the inability to monitor NCS 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 NCS. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1. If the NCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. MNS Basis Reference(s):

1. AP/1(2)/A/5500/10 NC System Leakage Within the Capacity of Both NV Pumps 2. PT/1(2)/A/4150/0010 Identifying NC System Leakage 3. NEI 99-01 CA1 D-79 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases C -Cold Shotdown I Refueling System Malfunction 1 -NCS Level Initiating Condition:

Loss of NCS inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency NCS water level cannot be monitored for;:::: 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

• UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory

• Visual observation of unisolable NCS leakage

• Reactor Building Refueling Bridge Monitor 1EMF16 (2EMF3) reading > 9000 mR/hr (Mode 6)

• Erratic Source Range or Wide Range Flux Monitor indication Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-6

• NCDT

• PRT

• CFAEsump

• ND/NS sump

• RHT

• WOT

• WEFT

• SRST Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

Sumps/Tanks 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. D-80 Revision 16-5 November, 2016

---

Basis: ATTACHMENT 1 EAL Bases The lowest measurable NCS level is the elevation of the NCS hot leg mid-loop.

Therefore, NCS inventory loss relative to the NCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications.

Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the NCS in areas outside the containment that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 1, 2). In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase.

The dose rate due to this core shine should result in indications on installed area radiation monitors.

1EMF16 (2EMF3), Reactor Building Refueling Bridge Monitor is located in the containment in proximity to the reactor cavity and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations.

If this radiation monitor reaches and exceeds 9,000 mR/hr (90% of instrument scale), a loss of inventory with potential to uncover the core is likely to have occurred.

Radiation monitors 1EMF16 and 2EMF3 are only required to be operable in Mode 6. Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

This IC addresses a significant and prolonged loss of reactor vessel/NCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a NCS 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 NCS level cannot be restored, fuel damage is probable.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).

It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor NCS 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 NCS . This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown D-81 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 MNS Basis Reference(s):

1. AP/1(2)/N5500/10 NC System Leakage Within the Capacity of Both NV Pumps 2. PT/1 (2)/N4150/001 D Identifying NC System Leakage 3. NEI 99-01 CS1 Category:

C -Cold Shutdown I Refueling System Malfunction Subcategory: 1 -NCS Level Initiating Condition:

challenged Loss of NCS inventory affecting fuel clad integrity with containment EAL: CG1 .1 General Emergency NCS level cannot be monitored for;::: 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

• UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory

• Visual observation of UNISOLABLE NCS leakage

• Reactor Building Refueling Bridge Monitor 1 EMF16 (2EMF3) reading > 9,000 mR/hr

• Erratic Source Range or Wide Range Flux Monitor indication AND Any Containment Challenge indication, Table C-1 Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

D-82 Revision 16-5 November, 2016 Table C-1 ATTACHMENT 1 EAL Bases Table C-6 Sumps/Tanks

• NCDT

• PRT

• CFAEsump

• ND/NS sump

• RHT

• WDT

• WEFT

• SRST Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

• Containment hydrogen concentration

> 6%

• UNPLANNED rise in containment pressure Mode Applicability: 5 -Cold Shutdown, 6-Refueling Definition(s):

CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to MNS, Containment Closure is established when the requirements of PT/1 (2)/A/4200/002 Care met. UNISOLABLE

-An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: The lowest measurable NCS level is the elevation of the NCS hot leg mid-loop.

Therefore, NCS inventory loss relative to the NCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications.

Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems D-83 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases connected to the NCS in areas outside the containment that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 1, 2). 1EMF16 (2EMF3), Reactor Building Refueling Bridge Monitor is located in the containment in proximity to the reactor cavity and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling' operations.

If this radiation monitor reaches and exceeds 9,000 mR/hr (90% of instrument scale), a loss of inventory with potential to uncover the core is likely to have occurred.

Radiation monitors 1EMF16 and 2EMF3 are only required to be operable in Mode 6. Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

Three conditions are associated with a challenge to containment integrity:

• CONTAINMENT CLOSURE is not established (ref. 3).

• 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 mixture of dissolved gases in the containment.

However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the containment atmosphere is greater than 6% (upper limit of for operability of hydrogen recombiners) by volume in the presence of oxygen (>5%) (ref. 4).

• Any unplanned increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability.

Unplanned containment pressure increases indicates containment closure cannot be assured and the containment cannot be relied upon as a barrier to fission product release. 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 meltjng 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 NCS level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.

If CONTAINMENT CLOSURE is 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 D-84 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).

It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor NCS 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 theNCS. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

MNS Basis Reference(s):

1. AP/1(2)/A/5500/10 NC System Leakage Within the Capacity of Both NV Pumps 2. PT/1 (2)/A/4150/001 D Identifying NC System Leakage 3. PT/1 (2)/A/4200/002 C Containment Closure 4. CALC MCC-1552-08-00-0208 Emergency Procedure Setpoints

5. NEI 99-01 CG1 D-85 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 2 -Loss of Essential AC Power Initiating Condition:

Loss of all but one AC power source to essential buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table C-2, to essential 4160V buses 1(2)ETA and 1(2)ETB reduced to a single power source 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

trable C-2 AC Power Sources Offsite:

• ATC (Train A)

• SATA (Train A)

• ATD (Train B)

• SATB (Train B) Onsite:

• DIG 1 (2) A (Train A)

• DIG 1 (2) B (Train B) 5 -Cold Shutdown, 6 -Refueling, NM -No Mode D-86 Revision 16-5 November, 2016 Definition(s):

ATTACHMENT 1 EAL Bases SAFETY SYSTEM-A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1). The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). Each essential bus has a dedicated diesel generator (D/G 1 (2) A & DIG 1 (2) B) to supply an onsite essential source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event but is not credited as an AC power source by Technical Specifications (ref. 1 ). This cold condition EAL is equivalent to the hot condition EAL SA 1.1. 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 no mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional D-87 Revision 16-5 November, 2016 __ j ATTACHMENT 1 EAL Bases 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 essential bus. Some examples of this condition are presented below.

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

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

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. MNS Basis Reference(s):

1. UFSAR Section 8.0 Electric Power 2. AP/1 (2)/A/5500/07 Loss of Electrical Power 3. ECA-0.0 EP/1(2)/A/5000/ECA-O.O Loss of All AC Power 4. NEI 99-01 CU2 D-88 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: CA2.1 Alert ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 2 -Loss of Essential AC Power Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1 (2)ETB for ;:: 15 min. (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that

• time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 5 -Cold Shutdown, 6 -Refueling, NM -No Mode Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1 ). The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). Each essential bus has a dedicated diesel generator (D/G 1 (2) A & DIG 1 (2) B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event. Although it is not credited as an AC power source by Technical Specifications, it is a credited source with regards to this EAL provided it is aligned within the 15 classification criteria (ref. 1 ). This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1. D-89 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 no mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an essential bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or RS1. MNS Basis Reference(s):

1. UFSAR Section 8.0 Electric Power 2. AP/1 (2)/A/5500/07 Loss of Electrical Power 3. ECA-0.0 EP/1 (2)/A/5000/ECA-O.O Loss of All AC Power 4. NEI 99-01 CA2 D-90 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 3 -NCS Temperature UNPLANNED increase in NCS temperature CU3.1 Unusual Event UNPLANNED increase in NCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s}:

UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3). In the absence of reliable NCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should NCS level indication be subsequently lost. This IC addresses an UNPLANNED increase in NCS temperature above the Technical Specification cold shutdown temperature limitand represents a potential degradation of the level of safety of the plant. If the NCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator 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.

This EAL involves a loss of decay heat removal capability, or an addition of heat to the NCS 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 at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel D-91 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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.

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.

MNS Basis Reference(s):

1. MNS Technical Specifications Table 1.1-1 2. MNS UFSAR Section 7.0 Instrumentation and Controls 3. AP/1 (2)/A/5500/19 Loss of ND or ND System Leakage System 4. NEI 99-01 CU3 D-92 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: -------------------

-ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 3 -NCS Temperature UNPLANNED increase in NCS temperature 1 CU3.2 Unusual Event Loss of all NCS temperature and NCS level indication for ;;::: 15 min. (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 5 -Cold Shutdown, 6-Refueling Definition(s):

None Basis: Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3). NCS water level is normally monitored using various instruments including NC System narrow range and wide range monitors, RVLIS, NC System sightglass, tygon tube and Pressurizer level instruments (ref. 4). This EAL addresses the inability to determine NCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the NCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3. This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor NCS 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.

D-93 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

MNS Basis Reference(s):

I 1. MNS Technical Specifications Table 1.1-1 2. MNS UFSAR Section 7.0 Instrumentation and Controls 3. AP/1 (2)/A/5500/19 Loss of ND or ND System Leakage System 4. OP/1 (2)/A/61 OO/SD-20 Draining the NC System 5. NEI 99-01 CU3 D-94 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: CA3.1 Alert ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 3 -NCS Temperature Inability to maintain plant in cold shutdown UNPLANNED increase in NCS temperature to > 200°F for> Table C-3 duration (Notes 1, 9) OR UNPLANNED NCS pressure increase > 20 psig due to a loss of NCS cooling (this does not apply during water-solid plant conditions)

• Note 1: Note 9: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the NCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6. Table C-3: NCS Heat-up Duration Thresholds NCS Status Containment Closure Heat-up Duration Status Intact (but not reduced N/A 60 min.* inventory)

Not intact established 20 min.* OR At reduced inventory not established 0 min.

• If an NCS heat removal system is in operation within this time frame and NCS temperature is being reduced, the EAL is not applicable.

Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

CONTAINMENT CLOSURE -The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to MNS, Containment Closure is established when the requirements of PT/1 (2)/A/4200/002 Care met. D-95 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases UNPLANNED

-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3). A 20 psig RPV pressure increase can be read on various instruments during outage (1NCLP5122 and 5142) (ref. 4). In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the NCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6. This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the NCS 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 NCS Heat-up Duration Thresholds table addresses an increase in NCS temperature when CONTAINMENT CLOSURE is established but the NCS is not intact, or NCS inventory is reduced (e.g., mid-loop operation).

The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The NCS Heat-up Duration Thresholds table also addresses an increase in NCS temperature with the NCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact NCS 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 NCS temperature, the NCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).

This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. The NCS pressure increase threshold provides a pressure-based indication of NCS up in the absence of NCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS1 or RS1. D-96 Revision 16-5 November, 2016 MNS Basis Reference(s):

1. MNS Technical Specifications Table 1.1-1 ATTACHMENT 1 EAL Bases 2. MNS UFSAR Section 7.0 Instrumentation and Controls 3. 1 AP/1(2)/A/5500/19 Loss of ND or ND System Leakage System 4. MCC-1210.04-00-0040

5. NEI 99-01 CA3 D-97 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 4 -Loss of Vital DC Power Initiating Condition:

Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event < 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for ;:: 15 min. (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

None Basis: The 125 VDC electrical power system consists of two independent and redundant safety related Class 1 E DC electrical power subsystems (Train A or EVDA, and Train B or EVDD). Each subsystem consists of two channels of 125 VDC batteries (each battery 100% capacity), the associated battery charger(s) for each battery, and all the associated control equipment and interconnecting cabling. (ref. 1 ). The Train A and Train B DC electrical power subsystems provide the control power for its associated Class 1 E AC power load group, 4.16 kV switchgear, and 600 V load centers. The DC electrical power subsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses. (ref. 1 ). The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 voe (ref. 1, 2). This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7 .1. 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. D-98 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.

For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an , I emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R. MNS Basis Reference(s):

1. MNS Technical Specification 3.8.4 DC Sources -Operating Bases 2. AP/1 /N5500/15 Loss of Vital or Aux Control Power 3. MNS UFSAR Section 8.0 Electrical Power 4. NEI 99-01 CU4 D-99 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases C -Cold Shutdown I Refueling System Malfunction 5 -Loss of Communications Initiating Condition:

Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 ORO communication methods OR Loss of all Table C-4 NRC communication methods Table C-4 Communication Methods System Public Address Internal Telephones Onsite Radios DEM NET Offsite Radio System Commercial Telephones NRC Emergency Telephone System (ETS) Mode Applicability: 5 -Cold Shutdown, 6 -Refueling, NM -No Mode Definition(s):

None Basis: On site x x x ORO NRC x x x x x Onsite/offsite communications include one or more of the systems listed in Table C-4 (ref. 1 ). D-100 Revision 16-5 November, 2016

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Public Address System ATTACHMENT 1 EAL Bases The McGuire Nuclear Station public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations.

Plant-wide instructions are issued using the paging feature. Internal Telephone System The McGuire Nuclear Station PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code. On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, TSC and EOF. DEMNET DEMNET is the primary means of offsite communication.

This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode. Offsite Radio System A dedicated radio network can be used for communication with county and state warning points. Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy. The local service provider provides primary and secondary power for their lines at the Central Office. NRG Emergency Telephone System The NRG uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRG regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network. Telephones connected to this network are located in the McGuire Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRG Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1. D-101 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 NRG. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are the State, Gaston, Catawba, Iredell, Lincoln, Cabarrus and Mecklenburg County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRG of an emergency declaration.

MNS Basis Reference(s):

1. MNS Emergency Plan Section F Emergency Communications

2. MNS Emergency Plan Section B On-Site Emergency Organization.

3. NEI 99-01 CU5 D-102 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases -------------------

-C -Cold Shutdown I Refueling System Malfunction 6 -Hazardous Event Affecting Safety Systems Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table C-5 hazardous event AND EITHER:

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

• The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Mode Applicability:

fable C-5 Hazardous Events

• Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift Manager 5 -Cold Shutdown, 6 -Refueling Definition(s):

EXPLOSION

-A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events require a post-event inspection to determine if the attributes of an explosion are present. D-103 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE -Damage to a 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.

Basis:

• The significance of seismic events are discussed under EAL HU2.1 (ref. 1 ).

• Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).

• External flooding may be due to high lake level. MNS powerhouse yard elevation is 760 ft MSL. The administration building and yard are elevation 747 ft MSL. The maximum water level elevation at the site is 760.375 ft MFL (ref. 3, 4).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of 95 mph. (ref. 5). *

• Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire response procedure (ref. 5). D-104 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases

• An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL. 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. The first conditional 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. The second conditional 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. MNS Basis Reference(s):

1. RP/O/A/5000/007 Earthquake

2. AP/O/A/5500/030 Plant Flooding 3. UFSAR Section 2.1 Site Location 4 UFSAR Section 3.4 Water Level (Flood) Design 5. UFSAR Section 3.3.1 Wind Loadings 6. AP/O/A/5500/45 Plant Fire 7. NEI 99-01 CA6 D-105 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category H -Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. 1.Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant. 2.Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. 3.Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown 5.Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant. D-106 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 6.Control Room Evacuation Events that are indicative of loss of Control Room habitability.

If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

?.Emergency Coordinator Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary.

The EALs of this category provide the Emergency Coordinator/EOF Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Coordinator/EOF Director judgment.

D-107 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H-Hazards 1 -Security Confirmed SECURITY CONDITION or threat HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervision Mode Applicability:

All Definition(s):

SECURITY CONDITION

-Any security event as listed in the approved security contingency plan that constitutes a threaUcompromise to site security, threaUrisk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. HOSTILE ACTION -An act toward MNS 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 MNS. 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). Basis: Security Shift Supervision is the Shift Security Supervisor or Response Team Leader. These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for MNS (Safeguards) information (ref. 1 ). This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one.of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1. D-108 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. This threshold references the Security Shift Supervison because these are the individuals trained to confirm that a security event is occurring or has occurred.

Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

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 Duke Energy Physical Security Plan for MNS (ref. 1 ). Escalation of the emergency classification level would be via IC HA 1. MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. NEI 99-01 HU1 D-109 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H-Hazards 1 -Security Confirmed SECURITY CONDITION or threat HU1.2 Unusual Event Notification of a credible security threat directed at the site Mode Applicability:

All Definition(s):

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. Basis: Security Shift Supervision is the Shift Security Supervisor or Response Team Leader. These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for MNS (Safeguards) information (ref. 1). This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73. 71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. This threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the MNS Security Contingency Plan (ref. 1 ). D-110 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 Duke Energy Physical Security Plan for MNS (ref. 1 ). Escalation of the emergency classification level would be via IC HA1. MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. NEI 99-01 HU1 D-111 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H-Hazards: 1 -Security Confirmed SECURITY CONDITION or threat HU1.3 Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

All Definition(s):

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. Basis: Security Shift Supervision is the Shift Security Supervisor or Response Team Leader. These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for MNS (Safeguards) information (ref. 1 ). This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73. 71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. This threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.

The status and size of the plane may also be provided by NORAD through the D-112 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases NRG. Validation of the threat is performed in accordance with the MNS Security Contingency Plan (ref. 1 ). 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 Duke Energy Physical Security Plan for MNS (ref. 1 ). Escalation of the emergency classification level would be via IC HA 1. MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. NEI 99-01 HU1 D-113 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: HA1.1 Alert ATTACHMENT 1 EAL Bases H-Hazards 1 -Security HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward MNS 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 MNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA -Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.

Basis: Security Shift Supervision is the Shift Security Supervisor or Response Team Leader. These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for MNS (Safeguards) information (ref. 1 ). 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 the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). D-114 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time: and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This IC 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. This threshold 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. 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 Duke Energy Physical Security Plan for MNS (ref. 1 ). MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. NEI 99-01 HA1 D-115 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: HA1.2 Alert ATTACHMENT 1 EAL Bases H-Hazards 1 -Security HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward MNS 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 MNS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA -Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.

Basis: Security Shift Supervision is the Shift Security Supervisor or Response Team Leader. These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for MNS (Safeguards) information (ref. 1 ). 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 the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). D-116 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of re'adiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This IC 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. This threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that 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-specific security procedures (ref. 2). The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.

The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.

This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.

Security-sensitive information should be contained in non-public documents such as the Duke Energy Physical Security Plan for MNS (ref. 1 ). MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. NEI 99-01 HA1 D-117 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H-:.--Hazards 1 -Security Hostile Action within the Protected Area HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward MNS 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 MNS. 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). PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area. Basis: Security Shift Supervision is the Shift Security Supervisor or Response Team Leader. These individuals are the designated on-site personnel qualified and trained to confirm that a security*

event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Duke Energy Physical Security Plan for MNS (Safeguards) information (ref. 1 ). This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3). D-118 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This IC 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 HA 1. 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 Duke Energy Physical Security Plan for MNS (ref. 1 ). Escalation of the emergency classification level would be via IC HG1. MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. NEI 99-01 HS1 D-119 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases H-Hazards 1 -Security Initiating Condition:

Hostile Action resulting in loss of physical control of the facility EAL: HG1.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred:

Any of the following safety functions cannot be controlled or maintained

• Reactivity control

• Core cooling

• NCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward MNS 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 MNS. 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). 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 PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area. D-120 Revision 16-5 November, 2016 Basis: ATTACHMENT 1 EAL Bases This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.

It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3). 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. 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 Duke Energy Physical Security Plan for MNS (ref.1 ). MNS Basis Reference(s):

1. Duke Energy Physical Security Plan for MNS 2. AP/O/A/5500/47 Security Events 3. AP/O/A/5500/48 Extensive Damage Mitigation

4. AP/1 (2)/A/5500/17 Loss of Control Room 5. NEI 99-01 HG1 D-121 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 2 -Seismic Event Seismic event greater than OBE levels HU2.1 Unusual Event Seismic event> QBE as indicated by OBE EXCEEDED alarm on 1AD-13, E7 Mode Applicability:

All Definition(s):

None Basis: Ground motion acceleration of 0.08g horizontal or 0.0533g vertical is the Operating Basis Earthquake for MNS (ref. 1, 3). Five strong motion triaxial accelerographs are used to obtain seismic event data at the station site. The seismic instrumentation system also consists of a network control center (NCC), which is used for rapid interrogation of the accelerograph data and for data transfer to a dedicated system computer for subsequent data processing and analysis.

The time-history recorded at each accelerograph location can be analyzed to determine its corresponding peak acceleration values and to verify that site Operating Basis Earthquake (OBE) limits have not been exceeded.

Immediate control room alarm indication of an earthquake of 0.08 g horizontal or 0.533 g vertical

• or greater is annunciated through the system's network control center (NCC), following seismic trigger actuation by at least two accelerographs (ref. 2). RP/O/A/5700/007 Earthquake provides the guidance for determining if the QBE earthquake threshold is exceeded and any required response actions. (ref. 4) To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of MNS. Provide the analyst with the following MNS coordinates:

35° 25' 59" north latitude, 80° 56' 55" west longitude (ref. 5). Alternatively, near real-time seismic activity can be accessed via the NEIC website: http://earthquake.

usgs.govleqcenterl D-122 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases An additional method to rule out spurious activation of the seismic instrumentation is to download seismic recorders stored memory on the dedicated laptop computer located in the Control Room, Elevation 767 ft., behind 1 MC9. Such validation should not, however, preclude a timely emergency declaration based on receipt of OBE alarm. 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 related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections).

Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. MNS Basis Reference(s):

1. UFSAR Section 3.1 Conformance with General Design Criteria 2. UFSAR Section 3.7.4.2 Location and Description of Instrumentation

3. OP/1/A/6100/010N Annunciator Response for Panel 1AD-13 4. RP/O/A/5700/007 Earthquake

5. UFSAR section 2.1.1 Site Location 6. NEI 99-01 HU2 D-123 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Hazardous event HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

All Definition(s):

PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area. Basis: Response actions associated with a tornado onsite is provided in RP/O/A/5700/006 Natural Disasters (ref. 1 ). If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1. A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on I Cs in Recognition Categories R, F, Sor C. MNS Basis Reference(s):

1. RP/O/A/5700/006 Natural Disasters

2. NEI 99-01 HU3 D-124 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Hazardous event HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode Mode Applicability:

All Definition(s):

FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. / SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis: Areas susceptible to internal flooding are Turbine Building, Service Building and Auxiliary from the following systems: Condenser Circulating Water, Fire Protection, Nuclear and Conventional Service Water and Condensate Storage (ref.1 ). Refer to EAL CA6.1 for internal flooding affecting one or more SAFETY SYSTEM trains. This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.

D-125 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the_ current operating mode. Escalation of the emergency classification level would be based on I Cs in Recognition Categories R, F, sore. MNS Basis Reference(s):

1. AP/O/A/5500/44 Plant Flooding 2. NEI 99-01 HU3 D-126 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Hazardous event HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) Mode Applicability:

All Definition(s):

/MPEDE(D)

-Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area. Basis: As used here, the term "offsite" is meant to be areas external to the MNS PROTECTED AREA. This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on I Cs in Recognition Categories R, F, Sor C. MNS Basis Reference(s):

1. NEI 99-01 HU3 D-127 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Hazardous event HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

All Definition(s):

None Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.

Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on I Cs in Recognition Categories R, F, Sor C. MNS Basis Reference(s):

1. NEI 99-01 HU3 D-128 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 4-Fire FIRE potentially degrading the level of safety of the plant HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1 ):

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

• Receipt of multiple (more than 1) fire alarms or indications

• Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Fire Areas

• Containment

• Auxiliary Building

• Diesel Generator Rooms

• FWST

• Dog Houses

• Standby Shutdown Facility (SSF) Mode Applicability:

All Definition(s):

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

Basis: The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The aiarm D-129 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Those actions could include visual observation or evaluation of thermal detector or pressure indicator data. Table H-1 Fire Areas are based on MCS-1465.00-00-0022 Design Basis Specification for the Appendix R Safe Shutdown Analysis and AP/O/A/5500/45 Plant Fire. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2). 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. For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed.

Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. MNS Basis Reference(s):

1. MCS-1465.00-00-0022 Design Basis Specification for the Appendix R Safe Shutdown Analysis 2. AP/O/A/5500/45 Plant Fire 3. NEI 99-01 HU4 D-130 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 4-Fire . Initiating Condition:

FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: . All Definition(s):

Table H-1 Fire Areas

• Containment

• Auxiliary Building

• Diesel Generator Rooms

• FWST

• Dog Houses

• Standby Shutdown Facility (SSF) 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.

Basis: The 30 minute requirement begins upon receipt of a single valid fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it D-131 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases is not spurious, or by reports from the field. Those actions could include visual observation or evaluation of thermal detector or pressure indicator data. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1. Table H-1 Fire Areas are based on MCS-1465.00-00-0022 Design Basis Specification for the Appendix R Safe Shutdown Analysis and AP/O/N5500/45 Plant Fire. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1, 2). 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. 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 HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis-Related Requirements from Appendix R 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 D-132 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 this EAL, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. MNS Basis Reference(s):

1. MCS-1465.00-00-0022 Design Basis Specification for the Appendix R Safe Shutdown Analysis 2. AP/O/A/5500/45 Plant Fire 3. NEI 99-01 HU4 D-133 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 4-Fire Initiating Condition:

FIRE potentially degrading the level of safety o(the plant EAL: HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

All Definition(s):

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

PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. MNS Basis Reference(s):

1. NEI 99-01 HU4 D-134 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 4-Fire Initiating Condition:

FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:

All Definition(s):

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

PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded.

The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.

Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. MNS Basis Reference(s):

1. NEI 99-01 HU4 D-135 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: HAS.1 Alert ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 5 -Hazardous Gases Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted Table H-2 Safe Operation

& Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area

!Auxiliary 716' P/C, RHole, near 1Nl-185, Outside CAD 212 !Auxiliary 750' 800 (1EMXA} 803 (1ETA} 702 (Elec. Pene.) !Auxiliary 733' 722 (1EMXB-1}

705 (1ETB} Mode Applicability: 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

Unit 2 Room/Area thru CAD Door, FF59 820 (2EMXA) 805 (2ETA) 713 (Elec. Pene.) 724 (2EMXB-1) 716 (2ETB} Modes 4 3,4 3,4 3 3,4 3,4 IMPEDE(D)

-Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis: If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. The list of plant rooms or areas with entry-related mode applicability identified specify those D-136 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown.

Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included.

In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1 ). 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 does not require atmospheric sampling; it only requires the Emergency Coordinator's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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

• The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).

For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

• The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

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

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

• If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.

This D-137 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 I Cs. NOTE: IC HAS mode applicability has been limited to the applicable modes identified in Table H-2 Safe Operation

& Shutdown Rooms/Areas.

If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a corresponding change to Attachment 3 'Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases' and to IC HAS mode applicability is required.

MNS Basis Reference(s):

1. Attachment 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases 2. NEI 99-01 HAS D-138 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 6 -Control Room Evacuation Initiating Condition:

Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility (SSF) Mode Applicability:

All Definition(s):

None Basis: The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation.

Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2). Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1. 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. D-139 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. AP/1 (2)/A/5500/17 Loss of Control Room 2. MCS-1465.00-00-0022 Appendix R Safe Shutdown Analysis 3. NEI 99-01 HA6 D-140 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 6 -Control Room Evacuation Initiating.Condition:

Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility (SSF) AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1 ):

• Reactivity (Modes 1, 2 and 3 only)

• Core Cooling

• NCS heat removal Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling Definition(s):

None Basis: The Shift Manager determines if the Control Room is inoperable and requires evacuation.

Control Room in habitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2). 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 following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Coordinator judgment.

The Emergency Coordinator 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).

D-141 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC FG1 or CG1 MNS Basis Reference(s):

1. AP/1 (2)/A/5500/17 Loss of Control Room 2. MCS-1465.00-00-0022 Appendix R Safe Shutdown Analysis 3. NEI 99-01 HS6 D-142 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: HU7.1 ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 7 -Emergency Coordinator Judgment Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE Unusual Event Other conditions exist which in the judgment of the Emergency Coordinator/EDF Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Mode Applicability:

All Definition(s):

SAFETY SYSTEM-A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis: The Emergency Coordinator/EDF Director are the designated onsite individuals having the responsibility and authority for implementing the MNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator/EDF Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Coordinator/EDF Director, emergency response personnel are notified and instructed to report to their emergency response locations.

In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, D-143 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases but Plant Management is expected to manage the emergency response as soon as. available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1 ). This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator/EDF Director to fall under the emergency classification level description for an Unusual Event. MNS Basis Reference(s):

1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator

2. NEI 99-01 HU? D-144 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: HA7.1 Alert ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 7 -Emergency Coordinator Judgment Other conditions exist that in the judgment of the Emergency Coordinator/EDF Director warrant declaration of an Alert Other conditions exist which, in the judgment of the Emergency Coordinator/EOF Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward MNS 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 MNS. 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). Basis: The Emergency Coordinator/EOF Director are the designated onsite individuals having the responsibility and authority for implementing the MNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations.

In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible ranging responsibilities associated with managing a major emergency (ref.1 ). D-145 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator/EOF Director to fall under the emergency classification level description for an Alert. MNS Basis Reference(s):

1. MNS Emergency Plan section B On-Site Emergency Organization Section 8.2 Emergency Coordinator

2. NEI 99-01 HA7 D-146 Revision 16-5 November, 2016

---

-,category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 7 -Emergency Coordinator Judgment Other conditions existing that in the judgment of the Emergency Coordinator/EDF Director warrant declaration of a Site Area Emergency HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator/EDF Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward MNS 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 MNS. 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) SITE BOUNDARY -Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 Site Boundary/Exclusion Area Boundary Basis: The Emergency Coordinator/EDF Director are the designated onsite individuals having the responsibility and authority for implementing the MNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations.

In this manner, the individual usually in charge of activities in the Control Room is D-147 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1 ). This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator/EOF Director to fall under the emergency classification level description for a Site Area Emergency.

MNS Basis Reference{s):

1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator

2. NEI 99-01 HS? D-148 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: HG7.1 ATTACHMENT 1 EAL Bases H -Hazards and Other Conditions Affecting Plant Safety 7 -Emergency Coordinator Judgment Other conditions exist which in the judgment of the Emergency Coordinator/EOF Director warrant declaration of a General Emergency General Emergency Other conditions exist which in the judgment of the Emergency Coordinator/EOF Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area Mode Applicability:

All Definition{s):

HOSTILE ACTION-An act toward MNS 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 MNS. 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). IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The Emergency Coordinator/EDF Director are the designated onsite individuals having the responsibility and authority for implementing the MNS Emergency Response Plan. The Operations Shift Manager(SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations.

In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible ranging responsibilities associated with managing a major emergency (ref. 1 ). D-149 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are oelieved by the Emergency Coordinator/EOF Director to fall under the emergency classification level description for a General Emergency.

MNS Basis Reference(s):

1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator

2. NEI 99-01 HG? D-150 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category S -System Malfunction EAL Group: Hot Conditions (NCS temperature

> 200°F); EALs in this category are applicable only in one or more hot operating modes. Numerous system-related equipment failure events that warrant emergency classification have been identified in this category.

They may pose actual or potential threats to plant safety. The events of this category pertain to the following subcategories: 1 . Loss of Essential AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite sources for 4160 VAC essential buses. 2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of vital plant 125 VDC power sources. 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification.

Losses of indicators are in this subcategory.

4. NCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% -5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category.

However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. NCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive NCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, NCS and containment integrity.

D-151 Revision 16-5 November, 2016

6. RPS Failure ATTACHMENT 1 EAL Bases This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not a_chieve reactor shutdown.

If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, NCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Containment Isolation Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification.

9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.

D-152 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 1 -Loss of Essential AC Power Initiating Condition:

Loss of all offsite AC power capability to essential buses for 1,5 minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and 1 (2)ETB for;:: 15 min. (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

rrable S-1 AC Power Sources Dffsite:

• ATC (Train A)

• SATA (Train A)

• ATD (Train 8)

• SATB (Train 8) Jnsite:

• DIG 1 (2) A (Train A)

• DIG 1 (2) 8 (Train 8) 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: Basis: The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train 8) (ref. 1). The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KVl4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a D-153 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KVl4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). Each essential bus has a dedicated diesel generator (DIG 1 (2) A & DIG 1 (2) B) to supply an . onsite emergency source of power to safe shutdown loads in the event of a loss of the normal ' power source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event but is not credited as an AC power source by Technical Specifications (ref. 1 ). 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 essential 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 essential buses, whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off site power. Escalation of the emergency classification level would be via IC SA1. MNS Basis Reference(s):

1. UFSAR Section 8.0 Electric Power 2. APl1 (2)1A/5500107 Loss of Electrical Power 3. ECA-0.0 EPl1 (2)1A/5000IECA-O.O Loss of All AC Power 4. NEI 99-01 SU1 D-154 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: SA1.1 Alert ATTACHMENT 1 EAL Bases S -System Malfunction 1 -Loss of Emergency AC Power Loss of all but one AC power source to essential buses for 15 minutes or longer AC power capability, Table S-1, to essential 4160V buses 1 (2}ETA and 1 (2}ETB reduced to a single power source 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

rl'able S-1 AC Power Sources Dffsite:

• ATC (Train A)

• SATA (Train A)

• ATD (Train B)

• SATB (Train B) Dnsite:

• DIG 1 (2) A (Train A)

• DIG 1(2) B (Train B) 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

SAFETY SYSTEM-A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: D-155 Revision 16-5 November, 2016 Basis: ATTACHMENT 1 EAL Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

For emergency classification purposes, "capability" means that an AC power source is available to the essential buses, whether or not the buses are powered from it. The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ET A (Train A) and ETB (Train B) (ref. 1 ). The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). , Each essential bus has a dedicated diesel generator (D/G 1 (2) A & DIG 1 (2) B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event but is not credited as an AC power source by Technical Specifications (ref. 1 ). The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

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

D-156 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases

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

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source. Escalation of the emergency classification level would be via IC SS1. MNS Basis Reference(s):

1. UFSAR Section 8.0 Electric Power 2. AP/1 (2)/A/5500/07 Loss of Electrical Power 3. ECA-0.0 EP/1 (2)/A/5000/ECA-0.0 Loss of All AC Power 4. NEI 99-01 SA1 D-157 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 1 -Loss of Emergency AC Power Initiating Condit.ion:

Loss of all offsite power and all onsite AC power to essential buses for 15 minutes or longer EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability to essential 4160V buses 1 (2)ETA and 1 (2)ETB for;:: 15 min. (Note 1) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: This EAL is indicated by the loss of all offsite and onsite AC power capability (Table C-2) to 4160V essential buses ETA and ETB. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1 ). For emergency classification purposes, "capability" means that an AC power source is available to the essential buses, whether or not the buses are powered from it. The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KVl4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KVl4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). Each essential bus has a dedicated diesel generator (DIG 1 (2) A & DIG 1 (2) B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event. Although it is not credited as an AC power source by Technical Specifications, it is a credited source with regards to this EAL provided it is aligned within the 15 minute classification criteria (ref. 1 ). D-158 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power capability are lost. 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. Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1. MNS Basis Reference(s):

1. UFSAR Section 8.0 Electric Power 2. AP/1 (2)/A/5500/07 Loss of Electrical Power 3. ECA-0.0 EP/1 (2)/A/5000/ECA-O.O Loss of All AC Power 4. NEI 99-01 SS1 D-159 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category:

S -System Malfunction Subcategory: 1 -Loss of Essential AC Power Initiating Condition:

1 Prolonged loss of all offsite and all onsite AC power to essential buses EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power capability to essential 4160V buses 1 (2}ETA and 1(2}ETB AND EITHER:

• Restoration of at least one essential bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

• Core Cooling RED PATH conditions met Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 4160V emergency buses ETA and ETB either for greater then the MNS Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1) or that has resulted in indications of an actual loss of adequate core cooling. Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. (ref. 2). The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KVl4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KVl4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). Each essential bus has a dedicated diesel generator (DIG 1 (2) A & DIG 1 (2) 8) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). D-160 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event (ref. 3). Four hours is the station blackout coping time (ref 2). Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Coordinator judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers.

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2). Specifically, Core Cooling RED PATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200°F or subcooling is 0°F AND no NC pumps are on AND core exit T/Cs are reading greater than or equal to 700°F AND Reactor Vessel Lower Range level less than or equal to 39% (ref. 2). This IC addresses a prolonged loss of all power sources to AC essential 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 essential 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 essential 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. D-161 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. UFSAR Section 8.4.2 Station Blackout Duration 2. EP/1 (2)/A/5000/F-O Critical Safety Function Status Tress -Core Cooling 3. UFSAR Section 8.0 Electric Power 4. AP/1 (2)/A/5500/07 Loi:;s of Electrical Power 5. ECA-0.0 EP/1 (2)/A/5000/ECA-Q.O Loss of All AC Power 6. NEI 99-01 SG1 D-162 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 1 -Loss of Essential AC Power Initiating Condition:

longer Loss of all essential AC and vital DC power sources for 15 minutes or . EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power capability to essential 4160V buses 1 (2)ETA and 1 (2)ETB for ;:: 15 min. AND Loss of all 125 VDC power based on battery bus voltage indications

< 105 VDC on both vital DC buses EVDA and EVDD for;:: 15 min. (Note 1) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown D-163 Revision 16-5 November, 2016 Definition(s):

None Basis: ATTACHMENT 1 EAL Bases This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4160V emergency buses ETA and ETB for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi. The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KVl4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KVl4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). Each essential bus has a dedicated diesel generator (DIG 1 (2) A & DIG 1 (2) B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1 ). An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event (ref. 1 ). The 125 VDC electrical power system consists of two and redundant safety related Class 1 E DC electrical power subsystems (Train A or EVDA, and Train B or EVDD). Each subsystem consists of two channels of 125 VDC batteries (each battery 100% capacity), the associated battery charger(s) for each battery, and all the associated control equipment and interconnecting cabling. (ref. 1 ). The Train A and Train B DC electrical power subsystems provide the control power for its associated Class 1 E AC power load group, 4.16 kV switchgear, and 600 V load centers. The DC electrical power subsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses. (ref. 1, 3). The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 voe (ref. 1, 3). This IC addresses a concurrent and prolonged loss of both essential AC and Vital DC power. A loss of all essential 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 D-164 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases sustained loss of both essential AC and vital 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. MNS Basis Reference(s):

1. UFSAR Section 8.0 Electric Power 2. AP/1 (2)/A/5500/07 Loss of Electrical Power 3. AP/1 (2)/A/5500/15 Loss of Vital or Aux Control Power 4. ECA-0.0 EP/1 (2)/A/5000/ECA-O.O Loss of All AC Power 5. NEI 99-01 SGB D-165 Revision 16-5 November, 2016 Category:

• Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 2 -Loss of Vital DC Power ' Initiating Condition:

Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Loss of all 125 VDC power based on battery bus voltage indications

< 105 VDC on both vital DC buses EVDA and EVDD for;::: 15 min (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: The 125 VDC electrical power system consists of two independent and redundant safety related Class 1 E DC electrical power subsystems (Train A or EVDA, and Train B or EVDD). Each subsystem consists of two channels of 125 VDC batteries (each battery 100% capacity), the associated battery charger(s) for each battery, and all the associated control equipment and interconnecting cabling. (ref. 1 ). The Train A and Train B DC electrical power subsystems provide the control power for its associated Class 1 E AC power load group, 4.16 kV switchgear, and 600 V load centers. The DC electrical power subsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses. (ref. 1, 2). The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 voe (ref. 1, 2). This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs RG1, FG1 or SG1. D-166 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. UFSAR Section 8.0 Electric Power 2. AP/1 (2)/A/5500/15 Loss of Vital or Aux Control Power 3. NEI 99-01 888 D-167 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases $ -System Malfunction 3 -Loss of Control Room Indications Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for;:: 15 min. (Note 1) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

able S-2 Safety System Param

• Reactor power

• NCS level

• NCS pressure

• Core exit TIC temperature

• Level in at least one SIG

• Auxiliary feed flow in at least one SIG Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Operator Aid Computer (OAC), which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2). This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. D-168 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases 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 NRG 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 NCS 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 SA3. MNS Basis Reference(s):

1. UFSAR Section 7.5 Safety-Related Display Instrumentation

2. OP/1(2)/N6100/SD-2 Cooldown to 400 Degrees F 3. NEI 99-01 SU2 D-169 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 3 -Loss of Control Room Indications Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room 15 min. (Note 1) AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

Table S-2 Safety System Parameters

• Reactor power

• NCS level

• NCS pressure

• Core exit TIC temperature

• Level in at least one SIG

• Auxiliary or emergency feed flow Table S-3 Significant Transients

• Reactor trip

• Runback > 25% thermal power

• Electrical load rejection

> 25% electrical load

• Safety injection actuation 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown D-170 Revision 16-5 November, 2016 Definition(s):

ATTACHMENT 1 EAL Bases UNPLANNED

-A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Basis: SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Operator Aid Computer (OAC), which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2). Significant transients are listed in Table S-3 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load or SI injection actuations.

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 NCS 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. D-171 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via I Cs FS1 or IC RS1 MNS Basis Reference(s):

1. UFSAR Section 7 .5 Safety-Related Display Instrumentation

2. OP/1(2)/N6100/SD-2 Cooldown to 400 Degrees F 3. NEI 99-01 SA2 D-172 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 4 -NCS Activity Initiating Condition:

NCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event NCS activity>

any of the following Technical Specification 3.4.16 limits:

• Dose Equivalent 1-131 > 1.0 µCi/gm for> 48 hrs.

• Dose Equivalent 1-131 > 60 µCi/gm

• Dose Equivalent Xe-133 > 280 µCi/gm Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s}:

None Basis: The specific iodine activity is limited to ::;; 1.0 µCi/gm Dose Equivalent 1-131 for> 48 hrs. or ::;; 60 µCi/gm Dose Equivalent 1-131 instantaneous.

The specific Xe-133 activity is limited to ::;; 280 µCi/gm Dose Equivalent XE-133 (ref 1, 2). 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. MNS Basis Reference(s}:

1. MNS Technical Specifications section 3.4.16 RCS Specific Activity 2. MNS Technical Specifications section 3.4.16 RCS Specific Activity Bases 3. NEI 99-01 SU3 D-173 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases S -System Malfunction 5 -NCS Leakage NCS leakage for 15 minutes or longer SU5.1 Unusual Event NCS unidentified or pressure boundary leakage> 10 gpm 15 min. OR NCS identified leakage > 25 gpm 15 min. OR Leakage from the NCS to a location outside containment

> 25 gpm for 15 min. (Note 1) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: Identified leakage includes leakage such as that from pump seals or valve packing (except reactor coolant pump (NCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, 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 NCS leakage through a steam generator to the secondary system (primary to secondary leakage) (ref. 1 ). Unidentified leakage is all leakage (except NCP seal water injection or leakoff) that is not identified leakage (ref. 1 ). Pressure Boundary leakage is leakage (except primary to secondary leakage) through a nonisolable fault in an NCS component body, pipe wall, or vessel wall (ref. 1) NCS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as NCS to the Component Cooling Water (KC), or systems that directly see NCS pressure outside containment such as Chemical & Volume Control System (NV), Nuclear Sampling system (NM) and Residual Heat Removal (ND) system (when in the shutdown cooling mode). D-174 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases This IC addresses NCS leakage which may be a precursor to a more, significant event. In this case, NCS 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. I The first and second EAL conditions are focused on a loss of mass from the NCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).

The third condition addresses an NCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.

The leak rate values for each condition were selected because they are usually observable with normal Control Room indications.

Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).

The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the NCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.

An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via I Cs of Recognition Category R or F. MNS Basis Reference(s):

1. MNS Technical Specifications Definitions section 1.1 2. NEI 99-01 SU4 D-175 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 6 -RPS Failure Initiating Condition: , Automatic or manual trip fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power;:: 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8) Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 1-Power Operations Definitlon(s):

None Basis: The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) trip function.

A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1 ). Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable.

A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4). D-176 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/1 S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4). Following any automatic RPS trip signal, EP/1 (2)/A/5000/E-O (ref. 2) and EP/1 S.1 (ref. 3) prescribe insertion of redundant manual trip signals to back up the automatic RPS trip function and ensure reactor shutdown is achieved.

Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event (ref. 4). In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required.

The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reduce reactor power below 5%, the event escalates to the Alert under EAL SA6.1. If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions. If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. D-177 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the, reactor, or a concurrent plant conditi9n, 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 plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 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 (e.g., RPS setpoint testing), the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor trip

• and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

MNS Basis Reference(s):

1. MNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. EP/1(2)/A/5000/E-O Reactor Trip or Safety Injection

3. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees -Subcriticality

4. EP/1 (2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS

5. NEI 99-01 SUS D-178 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 6 -RPS Failure Initiating Condition:

Automatic or manual trip fails to shut down the reactor EAL: SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is successful in shutting down the reactor as indicated by reactor power< 5% (Note 8) Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 -Power Operations Definition(s):

None Basis: This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power< 5% ). (ref. 1 ). Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable.

A predictable post-trip response from a manual reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3 4). For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/1(2)/A/5000/FR-D-179 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on : turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4). If both subsequent automatic and subsequent manual reactor trip actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 5%) following a failure of an initial manual trip, the event escalates to an Alert under EAL SA6.1 This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually 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 plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 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 (e.g., RPS setpoint testing), the following classification guidance should be applied. D-180 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases

• If the signal causes a plant transient that should have included an automatic reactor trip and the RTS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

MNS Basis Reference(s):

1. MNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. EP/1 (2)/A/5000/E-O Reactor Trip or Safety Injection

3. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees -Subcriticality

4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS

5. NEI 99-01 SU5 D-181 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 2 -RPS Failure Initiating Condition:

Automatic or manual trip fails to .shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power ;::5% AND Manual trip actions taken at the reactor control console (manual reactor trip switches or turbine manual trip) are not successful in shutting down the reactor as indicated by reactor power;:: 5% (Note 8) Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 -Power Operations Definition(s):

None Basis: This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed.

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/1 S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4). 5% rated power is a minimum reading on the power range scale that indicates continued power production.

It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent D-182 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases subsequent core damage. Below 5%, plant response will be similar to that observed during a normal shutdown.

Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1 ). Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers).

Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or NCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

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

MNS Basis Reference(s):

1. MNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. EP/1 (2)/A/5000/E-O Reactor Trip or Safety Injection

3. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees -Subcriticality

4. EP/1 (2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS

5. NEI 99-01 SA5 D-183 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 2 -RPS Failure Initiating Condition:

Inability to shut down the reactor causing a challenge to core cooling or NCS heat removal EAL: SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power ;:::so/o AND All actions to shut down the reactor are not successful as indicated by reactor power ;:::so/o AND EITHER:

• Core Cooling RED PATH conditions met

• Heat Sink RED PATH conditions met Mode Applicability:

1 -Power Operations Definition(s):

None Basis: This EAL addresses the following:

• Any automatic reactor trip signal followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SA6.1 ), and

• Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and NCS barriers.

Reactor shutdown achieved by use of EP/1 (2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) are also credited as a successful manual D-184 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1, 4 ). 5% rated power is a minimum reading on the power range scale that indicates continued power production.

It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below 5%, plant response will be similar to that observed during a normal shutdown.

Nuclear instrumentation can be used to determine if reactor power is greater than 5% power (ref. 1, 4). Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2). Specifically, Core Cooling RED PATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200°F or subcooling is 0°F AND no NC pumps are on AND core exit T/Cs are reading greater than or equal to 700°F AND Reactor Vessel Lower Range level less than or equal to 39% (ref. 2). Indication of inability to adequately remove heat from the NCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 2). Specifically, Heat Sink RED PATH conditions exist if narrow range level in at least on steam generator is not greater than or equal to 11 % (32% ACC) and total feedwater flow to the intact steam generators is less than or equal to 450 gpm. (ref. 3). 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 NCS. 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 shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to 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. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees -Subcriticality

2. EP/1 (2)/A/5000/F-O Critical Safety Function Status Tress -Core Cooling 3. EP/1 (2)/A/5000/F-O Critical Safety Function Status Tress -Heat Sink 4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS

5. NEI 99-01 SSS D-185 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 7 -Loss of Communications Initiating Condition:

Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 ORO communication methods OR Loss of all Table S-4 NRG communication methods Table S-4 Communication Methods System Onsite Public Address >< Internal Telephones x Onsite Radios x DEM NET Offsite Radio System Commercial Telephones NRG Emergency Telephone System (ETS) Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None D-186 ORO x x NRC x x Revision 16-5 November, 2016 Basis: ATTACHMENT 1 EAL Bases Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1 ). Public Address System The McGuire Nuclear Station public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations.

Plant-wide instructions are issued using the paging feature. Internal Telephone System The McGuire Nuclear Station PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code. On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, TSC and EOF. DEMNET DEMNET is the primary means of offsite communication.

This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode. Offsite Radio System A dedicated radio network can be used for communication with county and state warning points. Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy. The local service provider provides primary and secondary power for their lines at the Central Office. NRG Emeraency Telephone System The NRG uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRG regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network. Telephones connected to this network are located in the McGuire Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRG Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

D-187 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases This EAL is the hot condition equivalent of the cold condition EAL CU5.1. 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 (e.g., use of non-plant, privately owned equipment, relaying of site information via individuals or multiple radio transmission points, individuals being sent to off site locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are the State, Gaston, Catawba, Iredell, Lincoln, Cabarrus and Mecklenburg County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

MNS Basis Reference(s):

1. MNS Emergency Plan Section F Emergency Communications

2. MNS Emergency Plan Section B On-Site Emergency Organization.

3. NEI 99-01 CU5 D-188 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 8 -Containment Failure Initiating Condition:

Failure to isolate containment or loss of containment pressure control. EAL: SU8.1 EITHER: Unusual Event Any penetration is not isolated within 15 min. of a VALID containment isolation signal (Note 1) OR Containment pressure > 3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF for ;::: 15 min. (Notes 1, 10) Note 1: The Emergency Coordinator/EDF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 10: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes. Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis: The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.

One full train of containment cooling operating per design is considered (ref. 1, 2):

• One train of Containment Air Return Fan System (VX-CARF), and

• One train of Containment Spray System (NS) Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment D-189 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the GARF with a 10-minute delay) within 15 minutes this EAL has been exceeded.

At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control. The Unusual Event threshold applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes. The Unusual Event threshold also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to 15 minutes. Without a single train of VX-CARF in service following actuation, the Unusual Event should be declared regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes. This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent chalienges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant. For the first condition, the containment isolation signal must be generated as the result on an normal/accident condition (e.g., a safety injection or high containment pressure);

a failure resulting from testing or maintenance does not warrant classification.

The determination of containment and penetration status -isolated or not isolated -should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.

The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner. This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or NCS fission product barriers.

MNS Basis Reference(s):

1. MNS Technical Specification 3.6.6 2. MNS Technical Specification 3.6.6 Bases 3. MNS Technical Specification 3.3.2 4. UFSAR Section 6.2 Containment Systems 5. NEI 99-01 SU? D-190 Revision 16-5 November, 2016 Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 9 -Hazardous Event Affecting Safety Systems Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode -EAL: SA9.1 Alert The occurrence of any Table S-5 hazardous event AND EITHER:

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

• The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Mode Applicability:

Table 5-5 Hazardous Events

• Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Shift Manager 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

EXPLOSION

-A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

D-191 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in th'e cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE -Damage to a 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.

Basis:

• The significance of seismic events are discussed under EAL HU2.1 (ref. 1 ).

• Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).

• External flooding may be due to high lake level. MNS powerhouse yard elevation is 760 ft MSL. The administration building and yard are elevation 747 ft MSL. The maximum water level elevation at the site is 760.375 ft MFL (ref. 3, 4).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of 95 mph. (ref. 5).

• Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire response procedure (ref. 5).

• An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL. 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. The first condition 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 D-192 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second condition 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. MNS Basis Reference(s):

1. RP/O/A/5700/007 Earthquake

2. AP/O/A/5500/030 Plant Flooding 3. UFSAR Section 2.1 Site Location 4 UFSAR Section 3.4 Water Level (Flood) Design 5. UFSAR Section 3.3.1 Wind Loadings 6. AP/O/A/5500/45 Plant Fire 7. NEI 99-01 SA9 D-193 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category E -Independent Spent Fuel Storage Installation CISFSll EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold) An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

The MNS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISFSI would be applicable to EALs HU1.1, HA1.1and HS1.1 Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs. D-194 Revision 16-5 November, 2016 Category:

Sub-category:

ATTACHMENT 1 EAL Bases E-ISFSI None Initiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask> any Table E-1 dose limit Table E-1 ISFSI Dose Limits NAC Magnastor NAC UMS Transnuclear (TN-32)

• 190 mrem/hr

• 100 mrem/hr

• 120 mrem/hr (gamma) or 20 mrem/hr (gamma) on the side (neutron+ (neutron) on top of the cask of the cask gamma) on the 340 mrem/hr (gamma) or 40 mrem/hr (excludes air side of the cask

• inlet/outlet ports) (neutron) on the sides of the radial neutron

• 100 mrem/hr shield

• 10 mrem/hr (neutron+

560 mrem/hr (gamma) or 280 mrem/hr (neutron) on the side gamma) on the top

• of the cask of the cask (neutron) on the side surfaces above (excludes air radial neutron shield region inlet/outlet ports)

• 200 mrem/hr (neutron+

• 220 mrem/hr (gamma) or 400 mrem/hr

• 900 mrem/hr gamma) at air (neutron) on the side surfaces below the (neutron +gamma) inlets and outlets radial neutron shield region on the top of the cask (excludes air inlet/outlet ports) Mode Applicability:

All Definition(s):

CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the MNS ISFSI, Confinement Boundary is defined as the Transportable Storage Canister (TSC) for TN, UMS and MAGNASTOR storage systems. D-195 Revision 16-5 November, 2016 Basis: ATTACHMENT 1 EAL Bases The MNS ISFSI utilizes three designs for dry spent fuel storage:

• The Transnuclear (TN) TN-32 dry spent fuel storage system

• The NAC-UMS dry spent fuel storage system

• The NAC-MAGNASTOR dry spent fuel storage system All systems consist of a Transportable Storage Canister (TSC) and concrete Vertical Concrete Cask (VCC). The TSC is the CONFINEMENT BOUNDARY for all systems. The TSC is welded/bolted and designed to provide confinement of all radionuclides under normal, normal, and accident conditions (ref. 1, 2, 3). Confinement boundary is defined as the barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the TSC. The values shown in Table E-1 represent 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification for radiation external to a loaded cask for each of the NAC-MAGNASTOR, NAC-UMS and TN designs. All Table E-1 ISFSI dose limits are based on surveys taken consistent with the locations specified in the associated Technical Specification (ref. 1, 2, 3). This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category RIC RU1, is used here to distinguish between non-emergency and emergency conditions.

The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSls are covered under ICs HU1 and HA1. D-196 Revision 16-5 November, 2016 MNS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. TN Generic Technical Specifications

2. NAC-UMS Certificate of Compliance

3. MAGNASTOR Technical Specifications and Design Features 4. NEI 99-01 E-HU1 D-197 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (NCS temperature

> 200°F); EALs in this category are applicable only in one or more hot operating modes. EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment.

This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

The primary fission product barriers are: A. Fuel Clad (FCl: The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System (NCS): The NCS Barrier includes the NCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment (CMT): 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.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or NCS Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

• The Fuel Clad Barrier and the NCS Barrier are weighted more heavily than the Containment Barrier. D-198 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases

• U'nusual Event ICs associated with NCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

• For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification.

For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.

• The fission product barrier thresholds specified within a scheme reflect plant-specific MNS design and operating characteristics.

• As used in this category, the term NCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of NCS mass to any location-inside the primary containment, an interfacing system, or outside of the primary containment.

The release of liquid or steam mass from the NCS due to the designed/expected operation of a relief valve is not considered to be NCS leakage.

• At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration.

For example, if the Fuel Clad and NCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity.

Alternatively, if both the Fuel Clad and NCS fission product barriers were potentially lost, the Emergency Coordinator/EDF Director would have more assurance that there was no immediate need to escalate to a General Emergency.

D-199 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: FA1.1 Alert ATTACHMENT 1 EAL Bases Fission Product Barrier Degradation N/A Any loss or any potential loss of either Fuel Clad or NCS Any loss OR any potential loss of either Fuel Clad or NCS (Table F-1) Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: Fuel Clad, NCS and Containment comprise the fission product barriers.

Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and NCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or NCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability.

Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or NCS barrier results in declaration of a Site Area Emergency under EAL FS 1.1 MNS Basis Reference(s):

1. NEI 99-01 FA1 D-200 Revision 16-5 November, 2016 Category:

Subcategory:

Initiating Condition:

EAL: ATTACHMENT 1 EAL Bases Fission Product Barrier Degradation N/A Loss or potential loss of any two barriers : FS1.1 Site Area Emergency Loss OR potential loss of any two barriers (Table F-1) Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: Fuel Clad, NCS and Containment comprise the fission product barriers.

Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

• One barrier loss and a second barrier loss (Le., loss -loss)

• One barrier loss and a second barrier potential loss (i.e., loss -potential loss)

• One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss) At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important.

For example, the existence of Fuel Clad and NCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification.

Alternatively, if both Fuel Clad and NCS potential loss thresholds existed, the Emergency Coordinator/EOF Director would have greater assurance that escalation to a General Emergency is less imminent.

MNS Basis Reference(s):

1. NEI 99-01 FS1 D-201 Revision 16-5 November, 2016 ATTACHMENT 1 EAL Bases Category:

Fission Product Barrier Degradation Subcategory:

N/A Initiating Condition:

Loss of any two barriers and loss or potential loss of third barrier EAL: FG1 .1 General Emergency Loss of any two barriers AND Loss OR potential loss of third barrier (Table F-1) Mode Applicability: 1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: Fuel Clad, NCS and Containment comprise the fission product barriers.

Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

• Loss of Fuel Clad, NCS and Containment barriers

• Loss of Fuel Clad and NCS barriers with potential loss of Containment barrier

• Loss of NCS and Containment barriers with potential loss of Fuel Clad barrier

• Loss of Fuel Clad and Containment barriers with potential loss of NCS barrier MNS Basis Reference(s):

1. NEI 99-01 FG1 D-202 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment).

The table is structured so that each of the. three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds.

The fission product barrier categories are: A. NCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation I NCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.

The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each Loss and Potential Loss column -beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.1, the third Containment barrier Potential Loss in Category C would be assigned "CMT P-Loss C.3," etc. If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds.

This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded.

If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost -even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three D-203 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and NCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, and .FA 1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the NCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B, ... , E. D-204 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (NCS) Barrier Category Loss Potential Loss Loss Potential Loss 1. An automatic or manual ECCS 1. Operation of a standby charging A (SI) actuation required by pump is required by EITHER: EITHER: . UNISOLABLE NCS leakage NCS or SG None None Tube . UNISOLABLE NCS . SG tube leakage leakage Leakage 2. Integrity-RED PATH conditions . SG tube RUPTURE met 1. Core Cooling-ORANGE PATH B conditions met 1. Heat Sink-RED PATH conditions

1. Core Cooling-RED PATH 2. Heat Sink-RED PATH conditions met Inadequate conditions met met None AND Heat Removal AND Heat sink is required Heat sink is required c 1. EMF51A/B >Table F-2 column CMT "FC Loss" 1. EMF51A/B >Table F-2 column None "NCS Loss" None Radiation

2. Dose equivalent 1-131 coolant /NCS activity > 300 µCi/gm Activity D CMT None None None None Integrity or Bypass E 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the the Emergency Coordinator/EOF Director that indicates loss of the Emergency Coordinator/EOF Director Emergency Coordinator/EOF Director Emergency Coordinator/EOF Director EC fuel clad barrier that indicates potential loss of the fuel that indicates loss of the NCS barrier hat indicates potential Joss of the NCS Judgment clad barrier barrier D-205 Containment (CMT) Barrier Loss 1. A leaking or RUPTURED SG is FAULTED outside of containment None None 1. Containment isolation is required EITHER: . Containment integrity has been lost based on Emergency Coordinator/EOF Director judgment . UNISOLABLE pathway from Containment to the environment exists 2. Indications of NCS leakage outside of containment

1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates loss of the containment barrier Potential Loss None 1. Core Cooling-RED PATH conditions met Restoration procedures not effective within 15 min. (Note 1) 1. EMF51A/B >Table F-2 column GMT Potential Loss" 1. Containment-RED Path conditions met 2. Containment hydrogen concentration

>6% 3. Containment pressure > 3 psig with EITHER a failure of both trains of NS OR failure of both trains ofVX-CARF 15 min. (Notes 1. 1 O) 1. Any condition in the opinion of the Emergency Coordinator/EOF Director hat indicates potential loss of the containment barrier Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

1. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:

D-206 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

1. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None D-207 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

11. Core Cooling-RED PATH conditions met Definition(s):

None Basis: Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery.

The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1 ). This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. EP/1 (2)/A/5000/FR-C.1 Response to Inadequate Core Cooling 3. EP/1 (2)/A/5000/FR-C.2 Response to Degraded Core Cooling 4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A D-208 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: 1 . Core Cooling-ORANGE Path conditions met Definition(s):

None Basis: Cr i tical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates indicates subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1 ). This reading indicates a reduction in reactor vessel water level sufficien t to allow the onset of heat-induced cladding damage. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. EP/1 (2)/A/5000/FR-C

.1 Response to Inadequate Core Cooling 3. EP/1 (2)/A/5000/FR-C.2 Response to Degraded Core Cooling 4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A D-209 R evision 16-5 November , 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

2. Heat Sink-RED Path conditions met Heat sink is required Definition(s):

None Basis: In combination with NCS Potential Loss 8.1, meeting this threshold results in a Site Area Emergency.

Critical Safety Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1 ). The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1 ). The phrase "and heat sink required" precludes the need for classification for conditions in which NCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Red. Step 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 350°F (347°F . ACC). If these conditions exist, Heat Sink is required.

Otherwise, the operator is to either return to the procedure and step in effect or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 2). This condition indicates an extreme challenge to the ability to remove NCS 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.

D-210 Revision 16-5 November, 2016 ATTACHMENT

.2 Fission Product Barrier Loss/Potential*

Loss Matrix and Bases MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. EP/1 (2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink 3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B D-211 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

C. GMT Radiation I NCS Activity Degradation Threat: Loss Threshold:

11. EMF51A/B >Table F-2 column "FC Loss" Table F-2 Containment Radiation

-R/hr (EMF51A & B) [Time After SID (Hrs.) NCS Loss FC Loss CMT Potential Loss 0-1 8.8 550 5500 1-2 8.4 400 4000 2-8 7.0 160 1600 >8 6.2 100 1000 Definition(s):

None Basis: The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF51A & B. EMF51 & Bare located inside containment.

The detector range is approximately 1 to 1 EB R/hr (logarithmic scale). Radiation Monitors EMF51A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1 ). The Table F-2 values, column FC Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF51A & B) response based on a LOCA, for periods of 1, 2, 8 and >8 hours after shutdown, no sprays and NCS pressure < 1600 psig with -2% fuel failure (ref. 1 ). The value is derived as follows: RP/O/A/5700/019 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, and 8 and >8 hours after shutdown without spray and NCS pressure < 1600 psig x 0.02 (rounded) (ref. 1 ). D-212 Revision 16-5 November, 2016 ATTACHMENT 2 ' Fission Product Barrier Loss/Potential Loss Matrix and Bases The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this cond.ition indicates that a significant amount of fuel clad d9mage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for NCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the NCS Barrier. Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

MNS Basis Reference(s):

1. RP/O/A/5700/019 Core Damage Assessment

2. NEI 99-01 CMT Radiation I RCS Activity Fuel Clad Loss 3.A D-213 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

C. CMT Radiation I NCS Activity Degradation Threat: Loss Threshold:

2. Dose equivalent 1-131 coolant activity>

300 µCi/gm Definition(s):

None Basis: Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

The threshold dose equivalent 1-131 concentration is well above that expected for iodine spikes and corresponds to about 2% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost. (ref. 1 ). This threshold indicates that NCS radioactivity concentration is greater than 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. There is no Potential Loss threshold associated with NCS Activity I Containment Radiation.

MNS Basis Reference(s):

1. RP/O/A/5700/019 Core Damage Assessment

2. NEI 99-01 CMT Radiation I RCS Activity Fuel Clad Loss 3.B D-214 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

C. CMT Radiation I NCS Activity Degradation Threat: Potential Loss Threshold:

D-215 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

D. GMT Integrity or Bypass Degradation Threat: Loss Threshold:

D-216 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

D. GMT Integrity or Bypass Degradation Threat:

Potential Loss Threshold:

D-217 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates loss of the Fuel Clad barrier Definition(s):

None Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator/EOF Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost MNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A D-218 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates potential loss of the Fuel Clad barrier Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

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

MNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A D-219 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SI) actuation required by EITHER:

• UNISOLABLE NCS leakage

• SG tube RUPTURE Definition(s):

UNJSOLABLE

-An open or breached system line that cannot be isolated, remotely or locally. RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Basis: ECCS (SI) actuation is caused by (ref. 1 ):

• Pressurizer pressure < 1845 psig

• Containment pressure > 1.0 psig This threshold is based on an UNISOLABLE NCS 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 NCS Barrier. This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE NCS 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 FAUL TED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/E-O Reactor Trip or Safety Injection

2. EP/1 (2)/A/5000/E-3 Steam Generator Tube Rupture 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A D-220 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Category:

Degradation Threat: Threshold:

Reactor Coolant System A. NCS or SG Tube Leakage Potential Loss 1. Operation of a standby charging pump is required by EITHER:

• UNISOLABLE NCS leakage

• SG tube leakage Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: The Chemical and Volume Control System (CVCS) includes two centrifugal charging pumps which take suction from the Volume Control Tank and return cooled, purified reactor coolant to the NCS. Normal charging flow is handled by one of the two charging pumps. Each charging pump is designed for a flow rate of 150 gpm. A second charging pump being required is indicative of a substantial NCS leak. (ref. 1 ). This threshold is based on an UNISOLABLE NCS 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 NCS 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. MNS Basis Reference(s):

1. UFSAR Section 9.3.4 Chemical and Volume Control System D-221 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A D-222 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

12. Integrity-RED path conditions met Definition(s):

None Basis: The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity

-RED path. CSFST NCS Integrity

-Red Path plant conditions and associated PTS Limit A indicates an extreme challenge to the safety function when plant parameters are to the right of the limit curve following excessive NCS cooldown under pressure (ref. -1, 2). This condition indicates an extreme challenge to the integrity of the NCS pressure boundary due to pressurized thermal shock -a transient that causes rapid NCS cooldown while the NCS is in Mode 3 or higher (i.e., hot and pressurized).

MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. EP/1(2)/A/5000/FR-P.1 Response to Imminent Pressurized Thermal Shock Condition

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B D-223 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

D-224 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. Heat Sink-RED path conditions met Heat sink is required Definition(s):

None Basis: In combination with FC Potential Loss B.2, meeting this threshold results in a Site Area Emergency.

Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1 ). The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1 ). The phrase "and heat sink required" precludes the need for classification for conditions in which NCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat r Sink-Red.

Step 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 350°F. If these conditions exist, Heat Sink is required.

Otherwise, the operator is to either return to the procedure and step in effect or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 1, 2) This condition indicates an extreme challenge to the ability to remove NCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the NCS 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.

D-225 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold B.2; both will be met. This condition warrants a Site Area Emergency declaration because inadequate NCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase NCS pressure to the point where mass will be lost from the system. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. EP/1(2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink 3. NEI 99-01 Inadequate Heat Removal NCS Loss 2.B D-226 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

C. GMT Radiation/

NCS Activity Degradation Threat: Loss Threshold:

1. EMF51A/B > Table F-2 column "NCS Loss" Table F-2 Containment Radiation

-R/hr (EMF51A & B) Time After SID (Hrs.) NCS Loss FC Loss CMT Potential Loss 0-1 8.8 550 5500 1-2 8.4 400 4000 2-8 7.0 160 1600 >8 6.2 100 1000 Definition(s):

N/A Basis: The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF51A & B. EMF51A & Bare located inside containment.

The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale). Radiation Monitors EMF51A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1 ). The value specified represents, based on core damage assessment procedure RP/O/A/5700/019 Figure 1, the expected containment high range radiation monitor (EMF51A & B) response based on a LOCA, for periods of 1, 2, 8 and >8 hours after shutdown with no fuel failure (ref. 1 ). The value is derived as follows: RP/O/A/5000/019 Figure 1 Containment Radiation Level vs. Time for NCS Release for periods of 1, 2, 8 and >8 hours after shutdown (rounded) (ref. 1 ). The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical D-227 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the NCS Barrier only. There is no Potential Loss threshold associated with NCS Activity I Containment Radiation.

MNS Basis Reference(s):

1. RP/O/A/5700/019 Core Damage Assessment

2. NEI 99-01 CMT Radiation I RCS Activity NCS Loss 3.A D-228 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

B. GMT Radiation/

NCS Activity Degradation Threat: Potential Loss Threshold:

D-229 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

D. GMT Integrity or Bypass Degradation Threat: Loss Threshold:

D-230 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

D-231 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator/EDF Director that indicates loss of the NCS barrier Definition(s):

None Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS Barrier is lost. MNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment NCS Loss 6.A D-232 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates potential loss of the NCS barrier Definition(s):

None Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to the inability to reach final safety acceptance criteria before completing all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

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

MNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment NCS Potential Loss 6.A D-233 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:.

1. A leaking or RUPTURED SG is FAULTED outside of containment Definition{s):

FAUL TED -The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED -The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Basis: This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment.

The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for NCS Barrier Potential Loss A.1 and Loss A.1, respectively.

This condition represents a bypass of the containment barrier. FAUL TED 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 FAUL TED definition) and the FAUL TED 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 FAUL TED for emergency classification purposes.

The FAUL TED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification.

Steam releases of this size are readily observable with normal Control Room indications.

The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., NCS activity values) and IC SU5 for the NCS barrier (i.e., NCS 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 D-234 Revision 16-5 November, 2016 ATTACHMENT 2 , Fission Product Barrier Loss/Potential Loss Matrix and Bases steam to the environment (and are thus similar to a FAUL TED condition).

The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold.

Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown.

Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs. The ECLs 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-5 Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires operation of a Site Area Emergency per (makeup) pump (NCS Bamer Potential Loss) FS 1_1 Alert per FA1 .1 Requires an automatic or manual ECCS (SI) s*t A E . (NCS 8 . L ) 1 e rea mergency per Al FA 1 actuation amer oss FS 1 _ 1 ert per 1. There is no Potential Loss threshold associated with NCS or SG Tube Leakage. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/E-O Reactor Trip or Safety Injection

2. EP/1 (2)/A/5000/E-3 Steam Generator Tube Rupture 3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A D-235 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

D-236 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Category:

Degradation Threat: Threshold:

Containment B. Inadequate Heat Removal Potential Loss 1. Core Cooling-RED path conditions met AND Restoration procedures not effective within 15 min. (Note 1) Definition(s):

None Basis: Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery.

The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1 ). The function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions.

The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3). A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated.

If core exit thermocouple (TC) readings are greater than 1,200°F (ref. 1 ), Fuel Clad barrier is also lost. 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 NCS 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 challenge of the Containment Barrier. The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing.

Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Coordinator should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is D-237 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. EP/1 (2)/A/5000/FR-C.1 Response to Inadequate Core Cooling 3. EP/1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling 4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A D-238 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

C. CMT Radiation/NCS Activity Degradation Threat: Loss Threshold:

None D-239 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

C. GMT Radiation/NCS Activity Degradation Threat: Potential Loss Threshold:

1. EMF51 A/B > Table F-2 column "GMT Potential Loss" rrable F-2 Containment Radiation

-R/hr (EMF51A & B) rrime After SID (Hrs.) NCS Loss FC Loss CMT Potential Loss 0-1 8.8 550 5500 1-2 8.4 400 4000 2-8 7.0 160 1600 >8 6.2 100 1000 Definition(s):

None Basis: The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF51A & B. EMF51A & Bare located inside containment.

The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale). Radiation Monitors EMF51A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1 ). The Table F-2 values, column GMT Potential Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF51A & B) response based on a LOCA, for periods of 1, 2, 8 and >8 hours after shutdown, no sprays and NCS pressure < 1600 psig with -20% fuel failure (ref. 1 ). The value is derived as follows: RP/O/A/5700/019 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, 8 and >8 hours after shutdown without spray and NCS pressure < 1600 psig x 0.20 (rounded) (ref. 1 ). 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 D-240 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and NCS 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 NCS 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 ECL to a General Emergency.

MNS Basis Reference(s):

1. RP/O/A/5700/019 Core Damage Assessment

2. NEI 99-01 CMT Radiation I RCS Activity Containment Potential Loss 3.A D-241 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. GMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required EITHER:

• Containment integrity has been lost based on EC judgment

• UNISOLABLE pathway from containment to the environment exists Definition(s):

UNISOLABLE

-An open or breached system line that cannot be isolated, remotely or locally. Basis: These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Threshold

-Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage).

Following the release of NCS 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 accidenfconditions, it is expected that the Emergency Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.). Refer to the middle piping run of Figure 1. Two simplified examples are provided.

One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure. Another example would be a loss or potential loss of the NCS barrier, and the simultaneous occurrence of two FAUL TED locations on a steam generator where one fault is located inside containment (e.g., on a steam orfeedwater line) and the other outside of containment.

In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

D-242 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Following the leakage of NCS 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. Second Threshold

-Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment.

As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage).

Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful).

There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment.

Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an NCP seal cooler is allowing radioactive material to enter the Auxiliary Building.

The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well. Following the leakage of NCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components.

Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated' using the Recognition Category R ICs. The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1. MNS Basis Reference(s):

1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A D-243 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases ' Barrier: Containment Category:

D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of NCS leakage outside of containment Definition{s):

None Basis: , ECA-1.2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment.

Potential NCS leak pathways outside containment include (ref. 1, 2):

• Residual Heat Removal (ND)

• Safety Injection (NI)

• Chemical & Volume Control (NV)

• NCP seals (NC)

• PZR/NCS Loop sample lines (NM) 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 NCS 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 NCS 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 NCS mass is being lost outside of the containment.

Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building.

Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well. To ensure proper escalation of the emergency classification, the NCS leakage outside of containment must be related to the mass loss that is causing the NCS Loss and/or Potential Loss threshold A.1 to be met. D-244 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases' MNS Basis Reference(s):

1. EP/1 (2)/A/5000/ECA-1.2 LOCA Outside Containment

2. EP/1(2)/A/5000/E-1 Loss of Reactor or Secondary Coolant 3. NEI 99-01 CMT Integrity or Bypass Containment Loss D-245 Revision 16-5 November, 2016 ATTACHMENT 2 F i ssion Product Barrier Loss/Potential Loss Matrix a n d Bases Figure 1: Containment Integrity or Bypass Examples 1-----------I I Inside Reac t or Building Dampe r ----------

--, I I Area ' ------' I :Process I . :Moni t o r I I -----------

RCP D-246 Revision 1 6-5 November , 20 1 6 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

11. Containment-RED Path conditions met Definition(s):

None Basis: Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 15 psig and represents an extreme challenge to safety function. (ref. 1 ). 15 psig is based on the containment design pressure (ref. 2). 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 NCS 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. MNS Basis Reference(s):

1. EP/1 (2)/A/5000/F-O Critical Safety Function Status Trees 2. UFSAR Section 6.2 Containment Systems 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A D-247 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration

> 6% Definition(s):

None Basis: Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump. (ref. 1 ). The lower limit of deflagration of hydrogen in air is > 6% and is the maximum concentration at which hydrogen igniters can be placed in service (ref. 2). To generate such levels of combustible gas, loss of the Fuel Clad and NCS barriers must have occurred.

With the Potential Loss of the containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.

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. MNS Basis Reference(s):

1. UFSAR Section 6.2 Containment Systems 2. EP/1 (2)/A/5000/FR-Z.4 Response to High Containment Hydrogen Concentration

3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B D-248 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. GMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF 15 min. (Notes 1, 10) Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 1 O: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes. Definition(s):

None Basis: The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.

One full train of containment cooling operating per design is considered (ref. 1, 2):

• One train of Containment Air Return Fan System (VX-CARF), and

• One train of Containment Spray System (NS) Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the GARF with a 10-minute delay) within 15 minutes a potential loss of containment exists. At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control. The potential loss of containment applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes. The potential loss of containment also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to 15 minutes. Without a single train of VX-CARF in service following actuation, the potential loss should be credited regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes. D-249 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner. MNS Basis Reference(s):

1. MNS Technical Specification 3.6.6 2. MNS Technical Specification 3.6.6 Bases 3. MNS Technical Specification 3.3.2 4. UFSAR Section 6.2 Containment Systems 5. NEI 99-01 GMT Integrity or Bypass Containment Potential Loss 4.C D-250 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Category:

Degradation Threat: Threshold:

Containment F. Emergency Coordinator Judgment Loss 1. Any condition in the opinion of the Emergency Coordinator/EDF Director that indicates loss of the Containment barrier Definition{s):

None Basis: The Emergency Coordinator judgment threshold*

addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost. MNS Basis Reference{s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A D-251 Revision 16-5 November, 2016 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

F. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator/EDF Director that indicates potential loss of the Containment barrier Definition(s):

None Basis: The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

• This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost. MNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A D-252 Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases Background NEI 99-01 Revision 6 I Cs AA3 and HAS prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent.

Specifically the Developers Notes for AA3 and HAS states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down and shutdown.

Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations).

In addition, the list should specify the plant mode(s) during which entry would be required for each room or area. The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HAS: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

D-253 Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases MNS Table R-2 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown:

MNS P'rocedure and Step* , Step Action OP/1&2/A/6100/003, Perform OP/1/A/6100/SD 1 Enclosure 4.2, (Prepare For Cooldown).

Step 3.8.4.1 OP/1&2/A/6100/003 Perform NC System degas , Enclosure 4.2, per OP/1/A/6100/SD-10 Step 3.8.6 (NC System, PRT and NCDT Degas). OP/1&2/A/6100/003, Open breakers on Enclosure 4.2, Transformer Cooling Steps 3.8.8.1 & Groups. 3.8.9.1 OP/1&2/A/6100/003, Perform Main Steam Enclosure 4.2, Safety Valve testing.

Step 3.8.14 OP/1&2/A/6100/003, Check transfer of Aux Enclosure 4.2, Steam from C htr Bleed to Step 3.8.17.1 Main Steam (Close 1SP-1 (Main Steam to 1A CF Pump Turb !sol) and 1 SP-2 (Main Steam to 1 B CF Pump Turb lsol).1AS-11

). OP/1&2/A/6100/003 Stop G HOT Pumps per , Enclosure 4.2, OP/1/B/6250/004 Step 3.8.21 (Feedwater Heater Vents, Drains, and Bleed System). OP/1&2/A/6100/003, Stop C HOT Pumps per Enclosure 4.2, OP/1 /B/6250/004 Step 3.8.23 (Feedwater Heater Vents, Drains, and Bleed System). OP/1&2/A/6100/003, rTransfer of Aux Steam to Enclosure 4.2, Unit 2 or Aux Electric Step 3.8.34 Boilers per OP/1/B/6250/007 B (Auxiliary Electric Boilers).

OP/1&2/A/6100/003, Close 1SP-1 (Main Steam Enclosure 4.2, 1A CF Pump Turb !sol) Step 3.12.7 and 1 SP-2 (Main Steam to 1 B CF Pump Turb lsol). Building/Elevation/Room , , , ',*': N/A N/A rrransfer Yard Main Steam Doghouses rT°urbine Bldg. Basement (739') North Wall Turbine Bldg. Basement (739') West Wall Turbine Bldg. Basement (739') HP Heater Panel Service Bldg. (739') or V\uxiliary Boiler Room Turbine Bldg. Mezz (760') at CF Pumps D-254 Mode " N/A N/A 1 1 1 1 1 1 1 If action not performed does this prevent, ' cooldown/

No No No No No No No No No Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases ,. ***** *' ,. ,. MNS Procedure

• . .. Build.ing/Elevation/Room and Step ; : .. ' '* ' . . ' .. ". . OP/1&2/A/6100/003, Shutdown MG Sets per MG Set Room (767') Enclosure 4.2, OP/1/A/6150/008 (Rod Step 3.13.16.6 Control), Enclosure 4.5 (M/G Shutdown).

OP/1&2/A/6100/003, Secondary System Wet Secondary Chemistry Lab Enclosure 4.2, Layup Chemical addition and TB Basement (739') Step 3.14.4 (Chemistry).

OP/1&2/A/6100/003, Begin performance of Inside Containment Enclosure 4.2, stroke time testing of Pzr Step 3.14.6.2 PORVs. OP/1 &2/A/6100/S0-1, When RP allows access to Inside Containment Step 3.5 Lower Containment, begin Enclosure 4.2 (Pre-Cooldown Containment Entry). This enclosure performs a Containment Inspection with RP, Engineering and Operations involvement.

OP/1&2/A/6100/SO-After required amount of IAux. Bldg. (NM Lab 716') 1, Step 3.3.4 boron is added for SOM Counting Room (767') requirements for blocking P-11, Primary Chemistry samples NC System. OP/1&2/A/6100/SO-After required amount of Aux. Bldg. (NM Lab 716') 1, Step 3.4.9 boron is added for SOM Counting Room (767') Shutdown Boron Concentration, Primary Chemistry samples NC System. OP/1&2/A/6100/SO-After required amount of Aux. Bldg. (NM Lab 716') 1, Step 3.5.7 boron is added for Crud Counting Room (767') Burst Boron Concentration, Primary Chemistry samples NC System. OP/1&2/A/6100/SO-

!After required amount of Aux. Bldg. (NM Lab 716') 1, Step 3.6.7 boron is added for Counting Room (767') Refueling Boron Concentration, Primary Chemistry samples NC Svstem. ,, OP/1&2/A/6100/SO-Have Radwaste align Aux. Bldg. (716') D-255 ' Mode 3 3 3 3 3 3 3 3 1-3 If action not performed " does this p'revent :* cooldown/

shutdown?**

.. No No No No No No No No ' No Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases ' ' .. .*. '*!* ' MNS Procedure Step Action ' , Building{Elevatfon/Roo!TI and Step'r . ,,, '.o, .. ,, ' 10, Step 3.5.1.1 Nitrogen for NCDT Degas Radwaste Area per OP/1/A/6200/600 (WG Support of Unit 1 Shutdown).

OP/1&2/A/6100/SD-Radwaste performs Phase 1 Aux. Bldg. (716') 10, Step 3.6.2 PRT Degas per Radwaste Area OP/O/A/6200/518 (Waste Gas Operation).

OP/1&2/A/6100/SD-Radwaste performs NC lAux. Bldg. (716') 10, Step3.7.1 System Degas per Radwaste Area OP/1 &2/A/6200/600 (WG Support Of Unit 1/2 Shutdown).

OP/1&2/A/6100/SD-Radwaste performs NCDT lAux. Bldg. (716') 10, Step 3.8.3 Degas per OP/O/A/6200/518 Radwaste Area (Waste Gas Operation).

OP/1&2/A/6100/SD-Radwaste performs Phase 2 lAux. Bldg. (716') 10, Step 3.9.1 & PRT Degas per Radwaste Area 3.9.2 OP/1 &2/A/6200/600 (WG Support Of Unit 1/2 Shutdown) and OP/O/A/6200/518 (Waste Gas Operation).

,' ,, " " . .... ' .. . , .

: . '* ") . ., ,* < '* ".; **: " . . ,, "'""' OP/1&2/A/6100/SD-Radwaste crossties BATs. lAux. Bldg. (733') BAT 2, Step 3.2.3 lArea OP/1&2/A/6100/SD-When less than 1000 psig, Aux. Bldg. (733') Electrical 2, Step 3.5 IAE places NC System Pene (733' and 750') Narrow Range Pressure !Transmitters in service. OP/1&2/A/6100/SD-To maximize charging flow, Aux. Bldg. (733') Ledge 2, Enclosure 4.2, adjust NC Pump Seal Water Outside VCT Room Step 3.2.3.2 Injection Throttles to 8-10 gpm. ., '.' ***:* .* " e . '"' ,*' . '* . '* .** ,,, " .. ' *,' ' ' .; ' * '* ... ., OP/1&2/A/6100/SD-Radwaste crossties BATs. Aux. Bldg. (733') BAT Step 3.2.3 Area OP/1&2/A/6100/SD-Perform plant shutdown ETA, ETB, Aux. Bldg. 4, Steps 3.11. 1, (733' and 750') South End 3.11.2, & 3.12.3 OP/1&2/A/6100/SD-To maximize charging flow, Aux. Bldg. (733') Ledge Enclosure 4.2, adjust NC Pump Seal Water Outside VCT Room Step 3.10.2 Injection Throttles to 8-1 O gpm. D-256 Mode* '. 1-3 1-3 1-3 1-3 3 3 3 '" 3 3 3 If action not performed prevent u* '.>><

• shutdown?

• "i . No No No No .,,,. .,-.,, " No Yes No : No Yes No Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases .. MNS Procedure Step Action ;:ind Step. Building/Elevation/Ro-Om ' ... .. .* . OP/1&2/A/6100/SD-Secondary Chemistry to Chemistry Lab and 4, Step 3.15 check S/G sulfate meets Turbine Bldg. (739') North chemistry criteria for Wall cooldown.

I** ' '"'" .. :, I, OP/1&2/A/6100/SO-time testing of Inside Containment 10, Step 3.8 PORVs per PT/1/A/4151/005 (NC iValve Stroke Timing Test Using Air). Enclosures 13.4, 13.5, 13.6 OP/1&2/A/6100/SO-Rack out and tag one NV ET A (750'), ETB (733') 10, Step 3.11.1, land both NI Pumps per 3.11.2, 3.11.3 OP/O/A/6350/008 (Operation of Station Breakers).

OP/1&2/A/6100/SO-Tag out PD Pump at 1 MXK-Bldg. (750') North 10, Step 3.11.6 F2C (Reciprocating Charging End Pump No 1). OP/1&2/A/6100/SO-If L TOP vent requirements are Inside Containment 10, Step 3.12 o be satisfied by securing 1 NC-368 (Pzr PORV) open, Maintenance gags 1 NC-36B (Pzr PORV) . . . {* *.,. .. :*: ., .. *. . OP/1&2/A/6100/SD-Unlock and close 1/2ND-P/C, RHole, near 1Nl-185, 6A(B), Step 3.3 119 (1/2 A ND ECCS Outside CAD 212 (716') Sump Suction Relief Inlet ABPC thru CAD Door, lsol #2). FF59 (716') OP/1&2/A/6100/SD-Perform PT/1/A/4206/030 Aux Building Pipechase 6A(B), Enclosure (Draining ECCS Sump Piping (716') 14.1, Step 3.6 Drain Reservoir Train A), Enclosure 13.1 (Draining ECCS Sump Piping Drain Reservoir lrrain A in Modes 1 4) and PT/1/A/4206/031 (Draining ECCS Sump Piping Drain Reservoir Train B), Enclosure 13.1 (Draining ECCS Sump Piping Drain Reservoir Train B in Modes 1 4). OP/1&2/A/6100/SD-Monitor and shift NC System Aux Building (716'/733')

at 6A(B), Encl. 4 .. 1, Filters on high DP. NC Filters Room Step 3.8 D-257

• If action"riot performed Mode** does this prevent 4 ,, , 4 4 4 4 **' *.; 4 4 4

• cooldown/

shutdow.n? . '* .. No .. , "' .. , ., No Yes Yes No Yes No No Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases OP/1 &2/N6100/SD-6A{B), Encl. 4.2 Step 3.7 OP/1 &2/N61 OO/SD-6A{B), Step 3.31.1 De-energize 1/2ND-68A (A ETA Aux. Bldg. (750') ND Pump & A Hx Miniflow) in the open position at 1/2EMXA-F12B (1/2ND-68A). djust flow through Cation ux Building (750') over Bed Demineralizer per Demin Pits OP/1/N6200/001 D (Chemical and Volume Control System Dem ineralizers

). De-energize 1/2ND-67B (B ND Pump & B Hx Miniflow) in the open position at 1/2EMXB1-2C (1/2ND-67B). Table R-2 & H-2 Results 4 4 09s this cooldoy/nt*snutdown?J Yes No Table R-2/H-2 Safe Operation

& Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area 716' P/C,RHole, near 1Nl-185, Outside CAD 212 Auxiliary 750' 800 (1EMXA) 803 (1ETA) 702 (Elec. Pene.)

Auxiliary 733' 722 (1 EMXB-1) 705 (1ETB) Plant Operating Procedures Reviewed 1. OP/1&2/A/6100/003

2. OP/1&2/A/6100/SD-1

3. OP/1&2/A/6100/SD-2

4. OP/1&2/A/6100/SD-4

5. OP/1&2/A/6100/SD-10

6. OP/1 &2/A/61 OO/SD-6A(B)

7. OP/1&2/A/6100/S0-6

8. OP/1&2/A/6100/S0-10 Unit 2 Room/Area ABPC thru CAD Door, FF59 820 (2EMXA) 805 (2ETA) 713 (Elec. Pene.) 724 (2EMXB-1) 716 (2ETB) D-258 Modes 4 3,4 3,4 3 3,4 3,4 Revision 16-5 November, 2016 ATTACHMENT 3 Safe Operation

& Shutdown Areas Tables R-2 & H-2 Bases D-259 Revision 16-5 November, 2016 EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) 10 CFR 50.54(q) Screening Evaluation Form AD-EP-ALL-0602 Rev. 0 ATTACHMENT 4 Page 1 of 5 .. Applicable Sites *. Screening and Evaluation Number . . . . BNP D EREG #: 2076764 CNS D CR3 D HNP D MNS x 5AD #: 2069899 ONS D RNP D GO D Document and Revision MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-5 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan): ATIACHMENT2 F -Fission Product Barriers EALs FGl.1 and FSl.1 as pertaining to (Table F-1) Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Containment (CMT) Barrier Potential Loss Containment pressure>

3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF for> 15 min. (Note 1) changed to Containment pressure>

3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF for 15 min. (Notes 1, 10) Part II. Activity Previously Reviewed?

I Yes I I No Ix EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) Rev.a ATTACHMENT 4 Page 2 of 5 10 CFR 50.54(q) Screening Evaluation Form Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4 Evaluation is not , 10 CFR If yes, identify bounding source document number or approval reference and required.

Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part Ill Justification:

Attachment 4, Part V. Bounding document attached (optional) 10 *', '*. " . ;_ .... -' "" :* ,' ,, . .. ' . . . ' ' ... ** .. .' ' ' ' ". :..'-; ' ,, '" ' ' ' ., *' -_. ..

\ :._ *,*,-*.:*::

., . ,.*, ., *- "* Part Ill. Editorial Change Yes I No I x Is this activity an editorial or typographical change only, such as formatting, 10 CFR 50.54(q) Continue to paragraph numbering, spelling, or punctuation that does not change intent? Effectiveness Attachment 4, Evaluation is not Part IV and Justification:

required.

Enter address non justification and editorial complete changes Attachment 4, Part V & VI. Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)

Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box. -10, CFR so:41(b)(1)

Assignment of Responsibi,lity (OrganiZ(itiOn Control)<>

' *" ",, " ,., ,..,.** 1 ',. >>;.: ',' " -* ., " 1a Responsibility for emergency response is assigned.

D 1b The response organization has the staff to respond and to augment staff on a continuing basis D (24-7 staffing) in accordance with the emergency plan.

• 1 ci CFR

)(2) Onsite Emergency Organization . " \**, '< .-: .. '.,*,'.. 2 *' "" ".:: .. *: ., _:;;\* '.. .. 2a Process ensures that onshift emergency response responsibilities are staffed and assigned D 2b The process for timely augmentation of onshift staff is established and maintained.

D 3 10 CFR 5d.47(b}(3).Emergency Response Support and Resources

• * * ,., *' 3a Arrangements for requesting and using off site assistance have been made. D 3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. D (NA for CR3} 4 10 CFR 50.47(b)(4)

Emergency Classification System ,' . ' 4a A standard scheme of emergency classification and action levels is in use. x (Requires fina! approval of Screen and Evaluation by EP CFAM.) 5 10 CFR 50.47(b)(5)

Notification Methods and Procedures Sa Procedures for notification of State and local governmental agencies are capable of initiating notification D of the declared emergency within 15 minutes (30 minutes for CR3} after declaration of an emergency and providing follow-up notification.

5b 5c EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) 10 CFR 50.54{q) Screening Evaluation Form AD-EP-ALL-0602 Rev.a ATTACHMENT 4 Page 3 of 5 Administrative and physical means have been established for alerting and providing prompt instructions D to the public within the plume exposure pathway. (NA for CR3) The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and D Notification Systems for Nuclear Power Plants, or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. (NA for CR3)

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) 10 CFR 50.54(q) Screening Evaluation Form AD-EP-ALL-0602 Rev.a ATTACHMENT 4 Page 4 of 5 Part IV. Emergency Planning Element and Function Screen (cont.) . . 1 o CFR 50.47(b)(6)

Emergericy Comrm.mications

• :'.,, ', ... " ,*, ,*," " "' 6a Systems are established for prompt communication among principal emergency response D organizations.

6b Systems are established for prompt communication to emergency response personnel.

D 7)*,:* 10CFR 50.47(b)(7)

Publie; Education and Information,,*

• '*' .. ;;.*'* *.,* ' *.:\*., '., ' ' ' '" ' , *.' 7a Emergency preparedness information is made available to the public on a periodic basis within the D plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.

D ,8 1 O CFR$0.47(b)(8)

Emergency Faciiities arid Equipment

.*** ' ,, '" Ba Adequate facilities are maintained to support emergency response.

D 8b Adequate equipment is maintained to support emergency response.

D 9 10 CFR 50 .4 7(b )(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use. D 10 10 CFR 50.47(b)(10)

Protective 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) D 10b Evacuation time estimates for the population located in the plume exposure pathway EPZ are available D to support the formulation of PARs and have been provided to State and local governmental authorities. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during emergencies, including D those for hostile action events. 10d Kl is available for implementation as a protective action recommendation in those jurisdictions that D chose to provide Kl to the public. 11 * ' 1CJ'CFR50.47(b}(11)

Expos,ure CoritroL .* ,,:". ' -.,_ ' .. '*' 11a The resources for controlling radiological exposures for emergency workers are established.

D 10 CFR 50.47(b)(12)

Medical and Public Health Support.'

' ' " 12 :.'. 12a Arrangements are made for medical services for contaminated, injured individuals.

D 13 10 CFR 50.47(b)(13)

Recovery Planning and Post-accident Operations

,, ,, 13a Plans for rec9very and reentry are developed.

D 14 10 CFR 50.47(b)(14)

Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas) D is established.

14b Drills, exercises, and training evolutions that provide performance opportunities to develop, maintain, D and demonstrate key skills are assessed via a formal critique process in order to identify weaknesses.

14c Identified weaknesses are corrected.

D 15 10 CFR 50.47(b)(15)

Emergency Response Training 15a Training is provided to emergency responders.

D Part IV. Emergency Planning Element and Function Screen (cont.) 16 10 CFR 50.47(b)(16)

Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.

D EMERGENCY PLAN CHANGE SCREENING AND. AD-E:P'.'ALL-0602 10 CFR.50.54(9) .

,, ..... ATIACHMENT 4: .. ******* 10 CFR Screening Evaluatlo11 Form Page 5 of5 . Planners responsible for emergency plan development and maintenance are properly trained> *

• 0 * . If no Part IV criteria are a 1 O CFR Effectiveness Eilaluallon Is not required:

then complete*, Attachment 4, 1 o CFR 50.54(q) Screening Evaluation Fermi. Part V;: Go lo Attachment 4; 1 o CFR 50;54(q):'

• Screening Evaluation Form, Part VI for instructions describing the NRC required 30 day submittal.

.. If any Attachment 4; 10 Evaluation Form, Part IV* criteria are lhell complete Attachment 10 CFR Evaluatlon.Form; PartVat'ld perform a 10 CFRS0.54(q)

< *

• x Effectiveness Evaluation; Shaded block requires finat. approvalof Screen and EValuation by EP CFAM; >* **** Part Vt NRC Emergency Plan and Implementing Procedure Submittal Actions * * *

• Create two EREG General

• * .. One for EP. to provide the 10 CFR S0.54(q) summary of the analy51s, or the completed*

10

.* . x ** lo Licensing, . . . .< ... * : *.. *. ' . * .*. * . * .... * * *** * . .* . . *** .* .* ** .. *. * .....* *

• One for Licensing to submit the 1 O CFR Information to the NRC within 30 days after the change .* ** *** ** Is putin effect * * * * * *

• QA RECORD AD-EP-ALL-0602 EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) Rev. 1 ATTACHMENT 5 Page 1of8 << 10 CFR 50.54(q) Effectiveness Evaluation Form>> **>**. ., .,,. *::-.: ' 'screening a11d 'Evaluatioii N1J'mber*"'

' ,\.:: ,. ' '*i' .*."" . " . Applicable Sites *.>: ."':" ' -'*" ,. ' ... '*'.'-" *. BNP D EREG #: 2076764 CNS D CR3 D HNP D MNS x 5AD #:2069899 ONS D RNP D GO D Document and Revision

• MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-5 *-" ... -*------*-*-------* *--*-** -*-*-.. -.. .. .. .. *-,,. * .. "" .'_; .. ..... -Part I. Description of Proposed Change: MNS Emergency Plan Section D -Emergency Classification System Attachment 2: F -Fission Product Barriers EALs FG 1.1 and FS 1.1 as pertaining to (Table F-1) Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Containment (CMT) Barrier Potential Loss Containment pressure>

3 psig with EITHER a failure of both trains of NS OR failure of both trains ofVX-CARF for> 15 min. (Note 1) changed to Containment pressure > 3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF 15 min. (Notes 1, 10) Added a reference to NOTE 10 -If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes. Attachment 6, 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL} and EAL Yes x Bases Validation and Verification (V&V) Form , is attached (required for IC or EAL change) No D .* Part II. Description and Review of Licensing Basis Affected by the Proposed Change:

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) << 10 CFR 50.54(q) Effectiveness Evaluation Form>> DESCRIPTION MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-4 AD-EP-All-0602 Rev. 1 ATTACHMENT 5 Page 2 of 7 The Technical BASES changed because the methodology for the development of Emergency Action Levels changed FROM: "Regulatory Guide 1.101, Rev. 3, August 1992, approved the guidance provided byNUMARC/NESP-007, Revision 2, as an alternative methodology for the development of Emergency Action Levels. McGuire Nuclear Site used the NUMARC guidance for the development of initiating conditions and emergency action levels." TO "An emergency action level scheme based on the Nuclear Energy Institute (NEI) 99-01, Revision 6 which has been endorsed by the NRC in a letter dated March 28, 2013 from Mark Thaggard (U.S. Nuclear Regulatory Commission) to Susan Perkins-Grew (Nuclear Energy Institute) titled U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, November 2012 (MLl 2346A463)." The NRC approved the license amendments for MNS to adopt an emergency action level scheme based on the Nuclear Energy Institute (NEI) 99-01, Revision 6 (Development of Emergency Action Levels for Non-Passive Reactors).

Date of Issuance, May 24 2016 RlS 2005-02, Clarifying the Process for Making Emergency Plan Changes, Rev. 1 states, 2) Emergency plan a) The document prepared and maintained by the licensee that identifies and describes the licensee's methods for maintaining emergency preparedness, and responding to emergencies.

An emergency plan includes the plan as originally approved by the NRC, and all subsequent changes made by the licensee with, and without, prior NRC review and approval under 10 CFR 50.54(q).

i) The licensee's emergency plan consists of: (1) The emergency plan as approved by the NRC via a Safety Evaluation Report (SE), or license amendment (LA) from the Office of Nuclear Reactor Regulation (NRR) or the Office of Federal and State Materials and Environmental Management Programs (FSME). (2) Any subsequent changes to the emergency plan explicitly reviewed by the NRC through an SE, or LA from NRR or FSME, and found to meet the applicable regulations.

(3) Any subsequent changes made by the licensee without NRC review and approval after the licensee concluded that the change(s) do not constitute a decrease in effectiveness under 10 CFR 50.54(q).

This review concludes that the proposed changes continue to meet the intent and requirements established in the documents listed above. The effectiveness of the Emergency Plans is maintained.

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q) << 10 CFR 50.54(q) Effectiveness Evaluation Form>> Part Ill. Description of How the Proposed Change Complies with Regulation and Commitments.

AD-EP-All-0602 Rev. 1 ATTACHMENT 5 Page 3 of 7 If the emergency plan, modified as proposed, no longer complies with planning standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR Part 50, then ensure the change is rejected, modified, or processed as an exemption request under 10 CFR 50.12, Specific Exemptions, rather than under 10 CFR 50.54( q): This is a change to the Fission Product Barriers EALs FG 1.1 and FS 1.1 as pertaining to (Table F-1) as described in the BASES for threshold for containment barrier. This change does NOT impact or change the intent of EALs FG 1.1 and FS 1.1 as described in MNS Emergency Plan Section. The justification of these changes are as described:

This is a correction to the Containment (CMT) Barrier for a Potential Loss in Table F-1 Fission Product Barrier Threshold Matrix for the time of2: 15 min for failures of NS or VX as described in the BASES for the effective threshold.

BASES states " The potential loss of containment also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a l 0 minute delay for greater than or equal to 15 minutes." This is also a correction to the Containment (CMT) Barrier for a Potential Loss in Table F-1 Fission Product Barrier Threshold Matrix for reference to Note 10 as described in the BASES for the effective threshold.

BASES states Containment pressure > 3 psig with EITHER a failure of both trains of NS OR failur_e 9f b_oth trains_ of VX-CARF for 2: 15 min. (Notes l, 10)." Note l 0 provides information on the operation of the VX System: Note l 0: If the loss of containment cooling threshold is exceeded due to loss of both trains ofVX-CARF, this EAL only applies if at least one train ofVX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes. Compliance with IO CFR 50.47(b) and 10 CFR 50, Appendix E Requirements:

These proposed changes do not affect the emergency planning function associated with IO CFR 50.47(b)(4), because these changes continue to ensure that a standard scheme of emergency classification and action levels is in use at MNS. The EAL Technical Basis provides a valid, logical and intelligible methodology for the determination of an emergency classification.

The proposed changes continue to ensure that there is a means available for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials, including emergency action levels that are to be used as criteria for determining the need for notification and participation oflocal and State agencies, the NRC, and other Federal agencies.

The proposed changes continue to ensure use of emergency action levels for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety and include EALs for hostile action that may adversely affect the plant. In addition, the change continues to ensure emergency action levels that are based on in-plant conditions, in addition to onsite and offsite monitoring.

These proposed changes continue to support the requirements described in 10 CFR 50, Appendix E.IV.B.l.

The proposed changes continue to ensure the capability to assess, classify, and declare an emergency condition within fifteen minutes after the availability of indications that an EAL has been exceeded and to declare the emergency as soon as possible following identification of the appropriate emergency classification level. These proposed changes continue to support the requirements described in I 0 CFR 50, Appendix E.IV.C.2.

Part IV. Description of Emergency Plan Planning Standards, Functions and Program Elements Affected by the Proposed Change (Address each function identified in Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV of associated Screen):

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) << 10 CFR 50.54(q) Effectiveness Evaluation Form>> 50.47(b)(4), Emergency Classification System AD-EP-ALL-0602 Rev. 1 ATTACHMENT 5 Page 4 of 7 Risk Significant Planning Standard:

A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

RSPS Function:

The emergency planning function associated with 10 CFR 50.47(b)(4) states:

• A standard scheme of emergency classification and action levels is in use. Supporting requirements which are described in 10 CFR 50, Appendix E Sections IV. B (in part) and IV. C (in part) state: JJ. Assessment Actions: 1. The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation oflocal and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health ind safety. The emergency action levels shalfbe based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.

By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRC. Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis. C. Activatio11 ofE111erge11cy Orga11izatio11:

2. By June 20, 2012, 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 action 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 provided that any delay in declaration does not deny the State and local authorities the opportunity to implement measures necessary to protect the public health and safety. Program Elements:

• An emergency classification and emergency action level scheme must be established by the licensee.

• The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant emergency procedures.

• The initiating conditions include all postulated accidents in the Final Safety Analysis Report (FSAR) for the nuclear facility.

NUREG0654 NUREG 0654 D. Emergency Classification System Planning Standard states "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures." Evaluation Criteria 1. An emergency classification and emergency action level scheme as set forth in Appendix 1 must be established by the licensee.

The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant emergency procedures.

The plan shall identify the parameter values and equipment status for Applicability and Cross EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) << 10 CFR 50.54(q)

Effectiveness Evaluation Form>> Reference to Plans Licensee State Local each emergency class. AD-EP-ALL-0602 Rev. 1 ATTACHMENT 5 Page 5 of 7 2. The initiating conditions shall include the example conditions found in Appendix 1 and all postulated accidents in the Final Safety Analysis Report (FSAR) for the nuclear 1'(1ot1c.

NUREG 654 APPENDIX 1 U. S. NUCLEAR REGULATORY COMMISSION EMERGENCY ACTION LEVEL GUIDELINES FOR NUCLEAR POWER PLANTS Purpose of the site area emergency declaration is to ( 1) assure that response centers are manned, (2) assure that monitoring teams are dispatched, (3) assure that personnel required for evacuation of near-site areas are at duty stations if situation becomes more serious, (4) provide consultation with offsite authorities, and (5) provide updates for the public through offsite authorities.

Purpose of the general emergency declaration is to (1) initiate predetermined protective actions for the public, {2) provide continuous assessment of information from licensee and offsite organization measurements, (3) initiate additional measures as indicated by actual or potential releases, ( 4) provide consultation with offsite authorities and (5) provide updates for the public through offsite authorities.

Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:

The proposed changes described in Revision 16-05 of the MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM, do not result in a reduction in effectiveness of facilities, response organizations, or response equipment.

The proposed changes do not reduce the effectiveness of The McGuire Emergency Plan. The technical bases found in Section D will continue to provide technical guidance and instructions for EAL interpretation, resulting in an improved capability to classify an emergency in a timely manner. The proposed changes do not reduce the effectiveness of the currently approved EAL technical basis. These changes continue to accomplish the intent of Section D to provide clear guidance when classifying an emergency.

They will not result in delays in the review of the EALs nor will they cause an emergency classification to be modified or delayed. These enhancements will not change the process used to classify emergencies.

The proposed changes continue to support the emergency planning functions associated with 10 CFR 50.47(b)(4) as implemented by Emergency Plan. These changes ensure that a standard scheme of emergency classification and action levels remains in use at MNS by providing an overall improvement to the MNS Emergency Preparedness Program and enhancing ERO readiness to support a classified emergency.

This results in an improved capability to ensure the health and safety of plant personnel and the general public. The proposed revision of the MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM, incorporates consistency in the new EAL scheme and without a reduction in effectiveness to public health and safety. These changes continue to provide additional assurance that the McGuire Emergency Response Organization has the ability and capability to:

• respond to an emergency;

• perform functions in a timely manner;

• effectively identify and take measures to ensure protection of the public health and safety; and

• effectively use response equipment and emergency response procedures.

The proposed changes in Revision 16-05 the MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM, enhance the ability of MNS to support a classified emergency, resulting in an improved capability to ensure health and safety of plant personnel and the general public. These changes continue to meet NRG requirements, as described in 10 CFR 50.47(b) and 10 CFR 50, Appendix E and are an overall improvement to the EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) AD-EP-ALL-0602 Rev. 1 ATTACHMENT 5 Page 6 of 7 << 10 CFR 50.54(q) Effectiveness Evaluation Form>> MNS Emergency Preparedness Program. This change provides a more conservative scheme for EALs FGl.1 and FSl.1. Supporting Information:

SUB Failure to isolate containment or loss of containment pressure control SUB.1 Unusual Event states Any penetration is not isolated within 15 min. of a VALID containment isolation signal (Note 1) OR Containment pressure>

3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF 15 min. (Notes 1, 10) BASES for containment pressure is the same as the Containment (CMT) Barrier for a Potential Loss in Table F-1 Fission Product Barrier Threshold Matrix. There are no other classification-s other than unu-sual event based solely upon containment conditions.

EALs FGl.1 and FSl.1 use containment barrier combined with either fuel clad barrier or NCS barrier for classifications.

Part VI. Evaluation Conclusion.

Answer the following questions about the proposed change. 1 Does the proposed change comply with 10 CFR 50.47(b) and 10 CFR 50 Appendix E? Yes x No D 2 Does the proposed change maintain the effectiveness of the emergency plan (i.e., no reduction in effectiveness)?

Yes x No D 3 Does the proposed change maintain the current Emergency Action Level (EAL) scheme? Yesx No D 4 Choose one of the following conclusions:

a The activity does continue to com ply with the requirements of 1 O CFR 50.4 7(b) and 10 CFR 50, Appendix E, and the activity does not constitute a reduction in effectiveness or change in the current Emergency Action Level (EAL) scheme. Therefore, the activity can be implemented without prior NRC approval.

b The activity does not continue to comply with the requirements of 10 CFR 50.47(b) or 10 CFR 50 Appendix E or the activity does constitute a reduction in effectiveness or EAL scheme change. Therefore, the activity cannot be implemented without prior NRC approval.

Part VII. Disposition of Proposed Change Requiring Prior NRC Approval x D Will the proposed change determined to require prior NRC approval be either revised or II Yes D I No D rejected?

If No, then initiate a License Amendment Request in accordance 10 CFR 50.90 and AD-LS-ALL-0002, Regulatory Correspondence, and include the tracking number:

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) << 10 CFR 50.54(q) Effectiveness EvaluationForm.>>.

AD-EP-ALL-0602 .Rev.1 ATTACHMENT 5 .. Page7of 7 ** Part VIII. Signatures:

EP CFAM Final Approval Is required for changes affecting risk significant planning standard 10 CFR 50.47(b)(4). . .

• Approver (EP Manage/) (Print): * * <JbJfN

• Date: . *. //dd*/6 If the proposed activity is a change to the E-Plan or implementing procedures, then creale two EREG General Assignments.

_ _ _ _ _

• _ _ _

• One for EP to provide the. 10 CFR 50;54(q) summary ofthe analysis, orthe completed 10 CFR 50.54(q);

'h!f\ to Licensing.

• *

• p.*

• Orie for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after lhe change . L.'* is put In effect. }'l QA RECORD AD-EP-ALL-0602 EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) Rev. 1 ATTACHMENT 6 Page 1 of 5 << 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL) and EAL Bases Validation and Verification (V&V) Form >> Screening or Evaluation Nui:nber:,

• Part I. Identification of ICs and EALs Affected by Proposed Change: MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-5 ATIACHMENT2 F -Fission Product Barriers EALs FGl.1 and FSl.1 as pertaining to (Table F-1) Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Containment (CMT) Barrier Potential Loss Containment pressure>

3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF for> 15 min. (Note 1) changed to Containment pressure>

3 psig with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF for 15 min. (Notes 1, 10) Added NOTE 10 If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes. ,_ ,, .. *' .. Part II. Determination of Validation Method by Site EP Manager: In-Plant Walkdown D Tabletop D Training D Other (Specify)Verification of LAR x Simulator D NA D EP Manager Name (Print): EP Manager Date: Kevin L. Murray //./. ,,,1, ..

I/-/ti-/' Part Ill. Validation. (Answers marked No require resolution)

Validation Question Yes No NA Resolution and Comments

• Readouts, alarms, indications.

de., available in the D D x Control Room? Monitor, gauge, etc., designations are correct? D D x Are correct units of measure displayed on the D D x monitor, gauge, etc.? All values are within instrumentation display range? D D x EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) AD-EP-ALL-0602 Rev. 1 ATTACHMENT 6 Page 2 of 5 << 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL) and EAL Bases Validation and Verification (V&V) Form >> Is instrument display finite enough to distinguish D D x between values? No miscellaneous issues were identified during x D D walkdown correct? ' *.' " ."*:.:' *.c;:* ,' '" :V*' **;* .. **.'}* ;*:. ' :'.::*, ' ,; '., ' ' ' ,, ';,:ti ..... :*.)*>.,_:

' * '.< "*, : *. : .,:". ***** : . }'.', *,;:,"* H ' " .; .. :: -./'. " ' ,,,::.", ... : .. -'.

... ::' .*;*" ,,!>* *,. Part IV. Verification (Answers marked No require resolution)

';' *' '" " ValidatJon Ot;iestion

::>* '" -*Yes***

NA

• Resolytion and*

• .. :: ,. -'. " '* '* ... -, ... " " ',, " Is the IC/EAL change easy to use and does it flow well? Is sequencing logical and correct? x D D Is it written to appropriate level of detail and unambiguous?

Is the IC/EAL Matrix legible and easy to use? x D D Are correct units of measure displayed on the D D x monitor, gauge, etc.? Part IV. Verification (cont.) (No answers require resolution))

! ., validation auestion ./ .

* ** '<: '* "' Yes* N_d: *NA .. : _ ..* ,. * -. Resolution and

':,* .... .. .\ ***** . ' . . ' ,,' *\' "' .,. ,, " ' ' ....... " ' '*' Instrumentation; Plant Computer System (PCS); and/or Plant Process Computer System (PPCS) points specified?

• Correct instrument?

• Correct units

• Adequate instrument range? D D x

• Display unit readable?

• Proper significant digits?

• Instrument number and noun name provided?

• Consistent with operations procedures?

References specified in EAL Technical Basis x D D current and updated and source documents for inputs have been identified and verified to be appropriate for use? Does the change avoid human performance x D D challenges, latent weaknesses, and human performance traps?

• No vague or missing critical detail(s) . = Decisions are not over-reliant on knowledQe for successful performance Modifications, Emergency Plan, EAL Technical x D D Basis, reference manual and procedure revisions, setpoint changes, software changes, training, etc. are appropriately scheduled to correspond to the EAL revision?

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) AD-EP-ALL-0602 Rev. 1 ATTACHMENT 6 Page 3 of 5 << 10 CFR 50.54(q} Initiating Condition (IC} and Emergency Action Level (EAL} and EAL Bases Validation and Verification (V&V} Form >> Are alarm setpoints equal to or below EAL 0 D x thresholds?

Do radiation monitor setpoints account for 0 D x background?

'* Part V. Comments:

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS10 CFR 50.54(0).

..

1 . . . * *. * *** *. ATTACHMENT6

• ... * * . ** .. * .. **.* * .*.

• Page4of5 * << 1 o CFR 50.54(q) Initiating Condition. (IC) and Emergency Action Level (EAL) and EAL * * * *

• Bases Validation and Verification (V&V) Form >> *

• Part VI. Completion Review and Approval Signatures** . Validation and Verification (Print Names) [N<itel}:

Ren!lfd o. Bullis Validation and Verification Signatures: .

.* ... . . . .. . . .. Date: lf/10*1* . 16. Site EP Manager Review (Print Name): Date: Kevin L Murray

• ense Holder Signature Dale:. Qualified Emergency Coordina1or (Print Name): Date: .

• Jam trt;j G f\ * .

• 1-1/1m/1/p

• ** Engineering Review (Pri Name} [Note2J:

.. [BNP, CRJ; HNP, RNP] Fin PNSCApproval (Print Name)* * [BNP; CR , HNP, RNP]PNSC Signature M/(

. . AD-EP-ALL-0602 EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) Rev. 1 ATTACHMENT 6 Page 5 of 5 << 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL) and EAL Bases Validation and Verification (V&V) Form >> ,. .. ,-,._ ,*, '* ,"c '* -. * " , " ,, " --, , ,, *,,-' : ,, ,, '*!' ,. " ' *-' =: Part VII. NRC Emergency Plan and Implementing Procedure Submittal Actions If the*proposed activity is a change to the E-Plan or implementing procedures, then create two EREG General Assignments.

• One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR50.54(q), to Licensing.

• One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect. Notes: 1. Validation and Verification can be performed by same individual but must be:

• Qualified in the subject matter

• Separate from the author of change

• A cross-discipline reviewer 2. [BNP, CR3, HNP, RNPJ System specific Engineering Review is required for EAL changes related to process equipment such as radiological instruments and environmental monitoring.

3. [BNP, HNP, RNP] PSA review is required for EAL changes to ensure any potential or actual impact to PSA calculations or assumptions are adequately addressed. (Not applicable to CR3) QA RECORD __ I

, .. ' EMERGENCY PLAN CHANGE SCREENING AND -EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) 1 o* CFR 50.54(q) Screening Evaluation Form EREG #: 2074831 5AD #: 2069899 \ Document and Revision BNP CNS CR3 HNP MNS 'oNs RNP GO AD-EP-ALL-0602 Rev.a ATTACHMENT 4 Page 1 of 5 -D D D D x D D D MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-5 Part 1.-Description of Activity Being Reviewed (event or action, or series of actions th_at may result in a change to the emergency plan or affect the implementation of the emergency plan): Replaced BASES for EALs with the NRC NEI 99-01rev06 Technical BASES Revision 1 Final in Section D of the Emergency Plan. The pdf document from NRC was converted into a WORD document with page numbers and revision number added to the bottom footer. 10CFRS0.54q performed per EREG #2069911.

When validating the page numbers for the WORD conversion, it was discovered that there was incorrect numbering in the table of contents.

This editorial change has the same revision number as the previously reviewed NRC approved change. Numbering change for the following:

Table of Contents 2.1 changed to 2.2 changed to 2.1 2.3 changed to 2.2 2.4 changed to 2.3 2.5 changed to 2.4. 2.6 changed to 2.5 2.7 changed to 2.6 This corresponds with Section 2 for pages 3 -8. 3.1 changed to 3.0 3.2 changed to 3.1 3.3 changed to 3.2 These in Section 3 were also changed for pages 9 -13.

, EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) AD-EP-ALL-0602 Rev.a ATTACHMENT 4 Page 2 of 5 10 CFR 50.54(q) Screening Evaluation Form 4.1 changed to 4.0 4.2 changed to 4.1 4.3 changed to 4.2 This corresponds with Section 4 for page 14. 7.1 changed to 7.0 Formatting changes made through out documents due to the conversion from pdf to WORD such as page numbers, section breaks, paragraph alignment, headers, etc ... Part 1'1. Activity Previously Reviewed?

Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or

• Alert *and Notification System Design Report? If yes, identify.bounding source document number or approval reference and ensure the basis for concluding the source document fully bounds the proposed change is documented below: Justification:

Is this activity an editorial or typographical change only, such as formatting, paragraph numbering, spelling, or punctuation that does not change intent? Justification:

This is an editorial change where the numbering of the Table of Contents and Section 3 were corrected.

Formatting cha_nges made through out documents due to the conversion from pdf to WORD such as page numbers, section breaks, paragraph alignment, headers, etc ... This change does not change the intent of Section D of the MNS Emergency Plan. No EALs were affected.

,. ,. ':. '",: 10 CFR 50.54(q) Effectiveness Evaluation is not required.

Enter justification below and complete Attachment 4, Part V. 10 CFR 50.54(q) Effectiveness Evaluation is not required.

Enter justification and complete Attachment 4, Part V & VI. Continue to Attachment 4 , 10 CFR 50.54(q) Screening Evaluation Form; Part Ill Continue to Attachment 4, Part IV and address non editorial changes Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)

Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box. 1 . * . 10 CFR. 50:.47(b)(1)

Assignment of Responsibility (Organization coritrolf

.. 1 a Responsibility for emergency response is assigned.

D

.-EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) Rev.a ATTACHMENT 4 Page 3 of 5 10 CFR 50.54(q) Screening Evaluation Form 1 b The response organization has the staff to respond and to augment staff on a continuing basis D (24-7 staffing) in accordance with the emergency plan. 2a Process ensures that onshift emergency response responsibilities are staffed and assigned 2b The process for timely augmentation of onshift staff is established and maintained.

3a Arrangements for requesting and using off site assistance have been made. 3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. (NA for CR3) 4a A standard scheme of emergency classification and action levels is in use. (Requires final approval of Screen and Evaluation by EP CFAM.) 10 CFR,'50.47(b)(S}Nptification and Procedures:

... . . . . . ' . *. *.* .. D D D D D Sa Procedures for notification of State and local governmental agencies are capable of initiating notification D of the declared emergency within 1 S minutes (30 minutes for CR3) after declaration of an emergency and providing follow-up notification.

Sb Administrative and physical means have been established for alerting and providing prompt instructions D to the public within the plume exposure pathway. (NA for CR3) Sc The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and D Notification Systems for Nuclear Power Plants, or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. (NA for CR3) -!

I EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) * -10 CFR 50.54(q) Screening Evaluation Form AD-EP-ALL-0602 Rev. 0 ATTACHMENT 4 Page 4 of 5 Part IV. Emergency Planning Element and Function Screen (cont.) a;_. 1 o. CFR5.0.47(b)(6)

Cqmmi.mications;( _* -*. -: -' ----. , " \ '-";. ' >; *: . "* .. ,. .,._ ... 6a Systems are established for prompt communication among principal emergency response D organizations.

6b Systems are established for prompt communication to emergency response personnel.

D '*{':::*--_-

-, 10-CFR50.47(b}(?)

Public-Educati<:>n*and lnformat.ioiT':_,_;

.. , _:*_-::--. . :(:,"*'. *. ; ... ' ----* .. :' .. ,' -. .,,_ ._, '.-. :: *. ;*_.; *::.,. .. *i " 'i .. -"". 7a Emergency preparedness information is made available to the public on a periodic basis within the D plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.

D :*s:--/-*:

10

)(8) Emergency Facilities and'Ecjuipment.

• < / ' -----i;.,/; -** ,< * ..

' -----':._:-'--, __ --*: .: .*, __ _.':_:.,._--

--Ba Adequate facilities are maintained to support emergency response.

D 8b Adequate equipment is maintained to support emergency response.

D .g':' _ '10 CFR50.4 7(b )(9)° AcCident Assessnieiit.

• t*. ;-,: : --. .. : ,: ' : .. ... ' '. " --*-_, ,* ' .. *:

.. '.,. " .,* . --'-*\. . . , ,.-, ,-: **-9a Methods, systems, and equipment for assessment of radioactive releases are in use. D 10 CFR so'.47(b)(10)

Protective Response*.-" "' '. 10 " 10a A range of public PARs is available for implementation during emergencies. (NA for CR3) D 10b *Evacuation time estimates for the population located in the plume exposure pathway EPZ are available D to support the for:_mulation of PARs and have been_ provided to State and local governmental authorities. (NA for CR3) 10c A range of protective actions is available for plant emergency workers during e1T1ergencies, including D those for hostile action events. 10d Kl is available for implementation as a protective action recommendation in those jurisdictions that D chose to provide Kl to the public. :. 10-CFR Sd.47(b){-Hj Radiological Exposure Confror::

- --: :*:*., :-. > \ ,*'

./. >-' <:::; .. *. -.:::-:<i/ " :

.. , ' r *' ' . *' '. . ' ' ' -** "' ** *' ,-. '* ' < : . .;,-:', -.' :* -..-' ,,, -*.:;::_*.*

-*' .: <;,' 11a The resources for controlling radiological exposures for emergency workers are established.

D 10, CFR' so:4 7(b )(fa) Medical and-Public, Health Support : ':, , .. / .. 'c,* "* _, 12' ,* ",*: . ,.. : .. *-," ' -,. ., <**.-r "*' 12a Arrangements are made for medical services for contaminated, injured individuals.

D 13' 1 o CFRS0;4l(b)(13)

Plannihg:and Po.sFaccidenf Operations.:;,\

-, *-,:.: . * " -., .: /. -... -, , ' -" "*.:*.*, .. 13a Plans for recovery and reentry are developed.

D 14*. . 10 CFR50.47(b)(t4)

Drills*and.Exercises. .y __ --' .. -**, ,' .--.. /./;::: " -. ':r ' :' **:* i ,,'. *-_, 14a A drill and exercise program (including radiological, medical, health physics and other program areas) D is established.

14b Drills, exercises, and training evolutions that provide performance opportunities to develop, maintain, D and demonstrate key skills are assessed via a formal critique process in order to identify weaknesses.

14c Identified weaknesses are corrected.

D 15 10CFR 50.47(b)(.15)

Response Trainfng *** 15a Training is provided to emergency responders.

J D /

EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) Rev.O ) ATTACHMENT 4 Page 5 of 5 10 CFR 50.54(q) Screening Evaluation Form Part IV. Emergency Planning Element and Function Screen (cont.) 16a Responsibility for emergency plan development and review is established.

Planners responsible for emergency plan development and maintenance are properly trained. PART IV. Conclusion If no Part IV criteria are checked, a 10 CFR 50.54(q) Effectiveness Evaluation is not required, then complete Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part V. Go to Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, P'art VI for instructions describing the NRC required 30 day submittal.

If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete Attachment4, 10 CFR 50.54(q) Screening Evaluation Form, PartVand perform a 10 CFR 50.54(q) Effectiveness Evaluation.

Shaded block requires final approval of Screen and Evaluation by EP CFAM.

Date: II ltof t. Reviewer Nam (Print): Renard 0. Burris D Approver (EP Mana.geJ Name (Print):

Date: /f' //J /&-11o Approver (CF AM, as required)

Name '.JJ Part VI. NRC Emergency Plan and Implementing Procedure Su.bmittal Actions Create tWo EREG General Assignments.

/fl ft

• for to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR 50.54(q),

to Licensing. ,

... /t,

• Licensing fo submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change 1s put in effect. QA RECORD EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q) AD-EP-ALL-0602 Rev. 0 ATTACHMENT 4 Page 1 of 5 10 CFR 50.54(q) Screening Evaluation Form '.,-* .. _. : : Applicable

., Screening and Evaiuation Number'* .. ... -* _*,;: .. :. -.. *, **"\ ':**:: *'.' ,. *," .' .. , ",' ., *, -. '*" .* 1', ..* *\*' *('.< . \. . <*,' '".' ' *, '.,, ,, 1 *** . '*' '* " .. ., ": .. :. ' .,,, ,, ... .. ' BNP D EREG #: 2075426 CNS D CR3 D HNP D MNS x SAD #: 2069899 ONS D RNP D GO D Document and Revision MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-5 Part I. Description of ACtivity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or aff_ect the implementation of the emergency plan): Replaced BASES for EALs with the NRC approved NEI 99-01 rev 06 Technical BASES Revision 1 Final in Section D of the Emergency Plan .. 10CFRS0.54q performed per EREG #2069911.

-* ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System *Category:

A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

Changed "S/G tube RUPTURE" to "S/G tube leakage" Definition(s):

Deleted "RUPTURE" This is NOT a change to an EAL, this is only a change to a BASES description .

• 1 '. ., Part II. Activity Previously Reviewed?

I Yes I I No 1x EMERGENCY PLAN CHANGE SCREENING AND AD-EP-ALL-0602 EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q) Rev.a ATTACHMENT 4 Page 2 of 5 10 CFR 50.54(q) Screening Evaluation Form Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or 10 CFR 50.54(q) Continue to Alert and Notification System Design Report? Effectiveness Attachment 4 Evaluation is not I 10 CFR If yes, identify bounding source document number or approval reference and required.

Enter 50.54(q) ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below: below and Evaluation complete Form, Part Ill Justification:

Attachment 4, Part V. Bounding document attached (optional)

D . ', ' .. , . ' ,, " ' .: .*:*_,-;_,/

,'" __ * " -* '..*>. , : .. ***-.. _,,,,. *.-' . ... ", " " . -.*,, .. , "" ::*--... .-. "' *,_,\ ' *" ;._., ,c) *" '*

Part Ill. Editorial Change Yes I No x Is this activity an editorial or typographical change only, such as formatting, 10 CFR 50.54(q) Continue to paragraph numbering, spelling, or punctuation that does not change intent? Effectiveness Attachment 4, Evaluation is not Part IV and Justification:

required.

Enter address non justification and editorial

---

* -------cornplete cha_nges Attachment 4, Part V & VI. *.,_. Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)

Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box. f-:" *Jo CFR50:4:7(b}(1}-Assignmenfof . *" "* ,_ .' *'. ',._, " *, " .. *,,' 1a Responsibility for emergency response is assigned.

D 1b The response organization has the staff to respond and to augment staff on a continuing basis D (24-7 staffing) in accordance with the emergency plan. 2*.' '.'10* CFR'50A7(b)(2)

EmergericYbrgahization;

... <:. '.: -.. _. :.: ,., ",::, ". -.,_. .,* *:; '" : ... ,*,, ,. ;', ;! " ", *.,: ,'. .. , :-.. , ', ,. 2a Process ensures that onshift emergency response responsibilities are staffed and assigned D 2b The process for timely augmentation of onshift staff is established and maintained.

D : 3 .. :

• 1o:CFR 50.47(b)(3)

Emergency Sup-port and Resources . . "' -' ' :, '\ .,*,_. I ' * " -3a Arrangements for requesting and using off site assistance have been made. D 3b State and local staff can be accommodated at the EOF in accordance with the emergency plan. D (NA for CR3) _10 CFR q0.47(b)(4)

Emergency Classification System -* '" ,., " 4* ,* '*. . ' 4a A standard scheme of emergency classification and action levels is in use. x (Requires final approval of Screen and Evaluation by EP CFAM.) 5' .10 CFR 50.47(b)(5)

Notification Metho_ds and Procedures

• .. , 5a Procedures for notification of State and local governmental agencies are capable of initiating notification D of the declared emergency within 15 minutes (30 minutes for CR3) after declaration of an emergency and providing follow-up notification.

Sb Sc EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0) 10 CFR 50.54(q) Screening Evaluation Form AD-EP-ALL-0602 Rev.a ATTACHMENT 4 Page 3 of 5 Administrative and physical means have been established for alerting and providing prompt instructions D to the public within the plume exposure pathway. (NA for CR3) The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and D Notification Systems for Nuclear Power Plants, or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. (NA for CR3)

EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)

• AD-EP-ALL-0602 Rev. 0 ATTACHMENT 4 Page 4 of 5 10 CFR 50.54(q) Screening Evaluation Form Part IV. Emergency Planning Element and Function Screen (cont.) s'..'

CFR5o)l7(b)(6)

Emergen¢y'coin'nrnriications *** .... * * : * .,. , ,, ,* ,,.,.. -* . \, 6a Systems are established for prompt communication among principal emergency response organizations.

6b Systems are established for prompt communication to emergency response personnel.

.. ' .-' D D " .: ' . . :* . ' ' /,/.:* .... 7a Emergency preparedness information is made available to the public on a periodic basis within the plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.

a .

• 1_1:>CFR 50.47(b)(8)

Erne'rger:icy facilities and Equipment.

• **. . , . .. .* .. : .... *

• Ba Adequate facilities are maintained to support emergency response.

8b Adequate equipment is maintained to support emergency response.

9

• 1 o CFR 50.47(b)(9)

Accident

,. , .. *.* .. * * ... * *.. * . .' 9a Methods, systems, and equipment for assessment of radioactive releases are in use. 10 10 CFR 50.47(b)(10)

Protective Response 1 Oa A range of public PARs is available for implementation during emergencies. (NA for CR3) D D j,:' D D D D 1 Ob Evacuation time estimates for the population located in the plume exposure pathway EPZ are available D to support the formulation of PARs and have been provided to State and local governmental authorities. (NA for CR3) 1 Oc A range of protective actions is available for plant emergency workers during emergencies, including D those for hostile action events.

• 1 Od Kl is available for implementation as a protective action recommendation in those jurisdictions that chose to provide Kl to the public. D .. " U 10 CFR50.47(b)(11)

Radiological Exposure Cont,rol * .. :.**.>; .i' * : .. , . :--*.. ' *'. ... , ...... ,, .... --.. :. "' 11 a The resources for controlling radiological exposures for emergency workers are established.

D 1z.: 1Q*CFR5o:47(b)(12)

Medical and Pu'bllc Health_'SuppotL

.. ' \ .. \ ' :> I' .* ' ' . *'" -c '**, 12a Arrangements are made for medical services for contaminated, injured individuals.

D 13 ' *. 1 O CFR 50.4 7(b)(13) Recover-Y anc;i Post-accidenfOperatioris*

• ". .. .. * * .**. ...... ,' . .. : ,,, . 13a Plans for recovery and reentry are developed.

D ' ' I '. ' ' ': " ' . ::, ' ' . . ' . . ,

• 14 ,* 10 CFR 50.47(b)(14)

Drills ;;ind Exercise?s

\ * ... -... *" '* .'!,' .* .. ,:::* "' '* **,. 14a A drill and exercise program (including radiological, medical, health physics and other program areas) D is established.

14b Drills, exercises, and training evolutions that provide performance opportunities to develop, maintain, D and demonstrate key skills are assessed via a formal critique process in order to identify weaknesses.

14c Identified weaknesses are corrected.

D 15 10 CFR 50.47(b)(15)

Emergency Resp'onse Training 15a Training is provided to emergency responders.

D EMERGENCY PLAN CHANGE SCREENING AND AD-:EP-ALL-0602 EFFECTIVENESS EVALUATIOF\JS 10 CFR 50.54(0)

'. 1o. CFR 50.54(q) _Screening Evaluation Form . ATIACHMENT4 Page5of5:

• *

• Part Emergency Planning Element and Function Screen (cont.) ;.;._, .

• 16a Responsibility for emergency plan development and review is estiablished; . a ..

• 16b Planners responsiblefor emergency pfan development and maintenance are properly trained.

• a PART IV. Conclusion

• . * ** * * * * .*. . . * . . * :* * * .. * * * * . If no Part_ IV criteria are checked, a 1 O CFR 50.54(q)

Evaluation is not required; then complete _

• Attachment 4 1 .1 o. CFR Screening E;valuation Form, part VJ Go tQ Attachment 4, 1 O CFR 50.54(q) Screening Evaluation Form, PartVI for instructions describing the NRC required 30 day sutimlttaL

• .: . . **.* .* .* ' : **. . .* . : . . . : ... *, . . If any Attachment 4; 1DCFll50.54(q)

Screening Evaluation Form, Part IVcriteria are checked; then .. :--Attachment4, 10 CFR50.54(q)_Screening Evaluation Form; PartVand perform a 10 CFR50.54(q)

• r-Effecliveness Evaluation.

S_haded block requires final approval of Screen and Ellaluatron by EP CFAM> . *: .... : .. --------.

___ -_ ,-__ .-, -___ -.. -.,,-... -,

... *:.:.: ... "*,-,** -* Date:*: *-*. 11{ coJ 1 c.. . . . . *. RenardO.Bunis_

Approver (EP Manager Name (Print): . Date:: * .. * . _// .. , *.*. ',. Part VI. NRC Emergency Piaf) and lmplemeriling Procedure SubmitlalActfons Create two EREG General Assignments.

.. * . . * . .

• One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the.completed 10 CFR SO.S4(q);

x to Licensing. . * * *

• One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect; * -x QA RECORD ATTACHMENT 5 Page 1 of 8 << 10 CFR 50.54(q) Effectivenes.s Evaluation Form>> ... *Applicable:

Sites * .. ' -Screening and Evaluation Number. -;: .. -,,. BNP EREG #: 2075426 CNS CR3 HNP MNS SAD #:2069899 ONS RNP GO Document and Revision MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-5 . . :.-*, . . .. *. *, )\ ** *' ::;, Part I. Description of Proposed Change:* ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Category:

Degradation Threat: Threshold:

Reactor Coolant System A. NCS or SG Tube Leakage Potential Loss Changed "SIG tube RUPTURE" to "SIG tube leakage" Definition(s):

Deleted "RUPTURE" Attachment 6, 1 O CFR 50.54(q) Initiating Condition

{IC) and Emergency Action Level (EAL) and EAL Bases Validation and Verification (V&V) Form , is attached (required for IC or EAL change) -***' . .. .. . . D D D D x D D D Yes x No D ..... '* .*** .. ... .... ,, . .. . \,>., *.*.* .. , ... Part II. Description and Review of Licensing Basis Affected by the Proposed Change: DESCRIPTION MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM rev 16-4 The Technical BASES changed because the methodology for the development of Emergency Action Levels changed FROM: . . "Regulatory Guide 1.101, Rev. 3, August 1992, approved the guidance provided by NUMARC/NESP-007, Revision 2, as an alternative methodology for the development of Emergency Action Levels. McGuire Nuclear Site used the NUMARC guidance for the development of initiating conditions and emergency action levels." TO "An emergency action level scheme based on the Nuclear Enerav Institute (NEI) 99-01, Revision 6 which has been endorsed by the NRC in a letter dated March 28, 2013 from Mark Thaggard (U.S. Nuclear Regulatory Commission) to Susan Perkins-Grew (Nuclear Energy Institute) titled U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, November 2012(ML1 2346A463)." The NRC approved the license amendments for MNS to adopt an emergency action level scheme based on the Nuclear Energy Institute (NEI) 99-01, Revision 6 (Development of Emergency Action Levels for Non-Passive Reactors).

Date of Issuance, May 24 2016 RIS 2005-02, Clarifying the Process for Making Emergency Plan Changes, Rev. 1 states, 2) Emergency plan a) The document prepared and maintained by the licensee that identifies and describes the licensee's methods for maintaining emergency preparedness, and responding to emergencies.

An emergency plan includes the plan as originally approved by the NRC, and all subsequent changes made by the licensee with, and without, prior NRC review and approval under 10 CFR 50.54(q).

i) The licensee's emergency plan consists of: (1) The emergency plan as approved by the NRC via a Safety Evaluation Report (SE), or license amendment (LA) from the Office of Nuclear Reactor Regulation (NRR) or the Office of Federal and State Materials and Environmental Management Programs (FSME). (2) Any subsequent changes to the emergency plan explicitly reviewed by the NRC through an SE, or LA from NRR or.FSME, and found to meet the applicable regulations.

-(3) Any subsequent changes made by the licensee without NRC review and approval after the *licensee concluded that the change(s) do not constitute a decrease in effectiveness under 1 O CFR 50.54(q).

This review concludes that the proposed changes continue to meet the intent and requirements established in the documents listed above. The effectiveness of the Emergency Plans is maintained.

• • , ... * '.o.*.,* .: .* .. ; .. : *' * *** ** '> *:. ,,.::*

v Part Ill. Description of How the Proposed Change Complies with Regulation and Commitments.

If the emergency plan, modified as proposed, no longer complies with planning standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR Part 50, then ensure the change is rejected, modified, or processed as an exemption request under 10 CFR 50.12, Specific Exemptions, rather than under 10 CFR 50.54(q):

The term "SIG tube' RUPTURE" was mistakenly placed into the threshold for NCS or SIG Tube Leakage Potential Loss. The definition "RUPTURE" states that 'The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection" which would be a NCS or SIG Tube Leakage Loss. The threshold requires the Operation of a standby charging pump to be placed in service to restore and maintain pressurizer level as described in the BASIS of the Fission Product Barrier. Table F-1 Fission Product Barrier Threshold Matrix for NCS or SIG Tube Leakage Potential Loss states" Operation of a standby charging pump is required by EITHER UNISOLABLE NCS leakage or SG tube leakage. Table F-1 is part of the EAL and is not affected by this editorial change. NEl99-01 rev 06 development guide states on page 97 that SG tube leakage is the potential loss threshold for NCS or SIG Tube Leakage Loss.

• Compliance with 10 CFR 50.47(b) and 10 CFR 50, Appendix E Requirements:

These proposed changes do not affect the emergency planning function associated with 10 CFR 50.47(b)(4), .because these changes continue to ensure that a standard scheme of emergency*classification and action levels is in use at MNS. The EAL Technical Basis provides a valid, logical and intelligible methodology for the determination of an emergency classification.

The proposed changes continue to ensure that there is a means available for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the NRC, and other Federal agencies.

The proposed changes continue to ensure use of emergency action levels for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety and include EALs for hostile action that may adversely affect the-plant.

In addition, tne change continues-to ensure emergency action levels that are based on in-plant conditions, in addition to onsite and offsite monitoring.

These proposed changes continue to support the requirements described in 10 CFR 50, Appendix E.IV.B.1.

< The proposed changes continue to ensure the capability to assess, classify, and declare an emergency condition within fifteen minutes after the availability of indications that an EAL has been exceeded and to declare the emergency as soon as possible following identification of the appropriate emergency classification level. These proposed changes continue to support the requirements described in 10 CFR 50, Appendix E.IV.C.2.

• -*.* ' ' ": .... ' . ;. *". *' .... , ... '*. :*** ' '" '" ., **"" < '; "* Part IV. Description of Emergency Plan Planning Standards, Functions and Program Elements Affected by the Proposed Change (Address each function identified in Attachment 4; 10 CFR 50.54(q) Screening Evaluation Form, Part IV of associated Screen): 50.47(b)(4), Emergency Classification System .Risk Significant Planning Standard:

A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

RSPS Function:

The emergency planning function associated with 10 CFR 50.47(b)(4) states:

• A standard scheme of emergency classification and action levels is in use. Supporting requirements which are described in 10 CFR 50, Appendix E Sections IV. B (in part) and IV. C (in part) state: / B. Assessment Actions: 1. The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based.on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.

By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRC. Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis. C. Activation of Emergency Organization:

2. By June 20, 2012, 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 action 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 provided that any delay in declaration does not deny the State and local authorities the opportunity to implement measures necessary to protect the public health and safety.

• Program Elements:

• An emergency classification and emergency action level scheme must be established by the licensee.

• The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant emergency procedures.

• The initiating*condition-s include all postulafed accidents in the FinarSafety Anarysis-Report (FSAR) for the nuclear facility.

NUREG 0654 NU REG 0654 D. Emergency Classification System Planning Standard states "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures." Evaluation Criteria 1. An emergency classification and emergency action level scheme as set forth in Appendix 1 must be established by the licensee.

The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant emergency procedures.

The plan shall identify the values and equipment status for Applicability and Cross Reference to Plans Licensee State Local each emergency class. 2. The initiating conditions shall include the example conditions found in Appendix 1 and all postulated accidents in the Final Safety Analysis Report (FSAR) for the nuclear facility.

""1\ ll (1<>( <IP

• NUREG 654 APPENDIX 1 (jb U. S. NUCLEAR REGULATORY COMMISSION EMERGENCY ACTION LEVEL GUIDELINES FOR NUCLEAR POWER PLANTS Purpose of the site area emergency declaration is to (1) assure that response centers are manned, (2) assure that monitoring teams are dispatched, (3) assure that personnel required for evacuation of near-site areas are at duty stations if situation becomes more serious, (4) provide consultation with offsite authorities, and (5) provide updates for the public through offsite authorities.

Purpose of the general emergency declaration is to (1) initiate predetermined protective actions for the public, {2) provide continuous assessment of information from licensee and offsite organization measurements, (3) initiate additional measures as indicated by actual or potential releases, (4) provide consultation with offsite authorities and (5) provide updates for the public through offsite authorities.

Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:

' . The proposed changes described in Revision 16-05 of the MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM, do not result in a reduction in effectiveness offacilities, response organizations, or response equipment.

The proposed changes do not reduce the effectiveness of The McGuire Emergency Plan. The technical bases found in Section D will continue to provide technical guidance and instructions for EAL interpretation, resulting in an improved capability to classify an emergency in a timely manner. The proposed changes do not reduce the effectiveness of the currently approved EAL basis. These changes continue to accomplish the intent of Section D to provide clear guidance when classifying an emergency.

They will not result in delays in the review of the EALs nor will they cause an emergency classification to be modified or delayed. These enhancements will not change the process used to classify emergencies.

The proposed changes continue to support the emergency planning functions associated with 10 CFR 50.47(b)(4) as implemented by Emergency Plan. These changes ensyre that a standard scheme of emergency classification and action levels remains in use at MNS by providing an overall improvement to the MNS Emergency Preparedness Program and enhancing ERO readiness to support a classified emergency.

This results in an improved capability to ensure the health and safety of plant personnel and the general public.

• The proposed revision of the MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM, incorporates consistency in the new EAL scheme and without a reduction in effectiveness to public health and safety. These changes continue to provide additional assurance that the McGuire Emergency Response Organization has the ability and capability to:

• respond to an emergency;

• *

• perform functions in a timely manner; and take_

t() ensure public he_alth_ and safet}'.;

and effectively use response equipment and emergency response procedures . The proposed changes in Revision 16-05 the MNS Emergency Plan Section D EMERGENCY CLASSIFICATION SYSTEM, enhance the ability of MNS to support a classified emergency, resulting in an improved capability to ensure health and safety of plant personnel and the general public. These changes continue to meet NRC requirements, as described in 10 CFR 50.47(b) and 10 CFR 50, Appendix E and are an overall improvement to the MNS Emergency Preparedness Program. ... -:* '-*;*:".*

.. -*'-'.: **:-* .. --.-..

• -

• . ' .. *., , -*:*: '* : . . . ' '*' ,_ ,. -* c.* '";'" .. -.::> .. ;/i* *.**. " '" **-*

,, ' " *: " , .. Part VI. Evaluation Conclusion.

Answer the following questions about the proposed change.* 1 . Does the proposed change comply with 10 CFR 50.47(b) and 10 CFR 50 Appendix E? Yesx No D 2 Does the proposed change maintain the effectiveness of the emergency plan (i.e., no reduction in effectiveness)?

Yesx No D .. 3 Does the proposed change maintain the current Emergency Action Leve_I (EAL) scheme? Yesx No D 4 Choose one of the following conclusions:

a The activity does continue to comply with the requirements of 10 CFR 50.47(b) and 10 CFR 50, Appendix E, and the activity does not constitute a reduction in effectiveness or change in the current Emergency Action Level (EAL) scheme. Therefore, the activity can be implemented without prior NRC approval.

b The activity does not continue to comply with the requirements of 10 CFR 50.47(b) or 10 CFR 50

• Appendix E or the activity does constitute a reduction in effectiveness or EAL scheme change. Therefore, the activity cannot be implemented without prior NRC approval.

Part VII. Disposition of Proposed Change Requiring Prior NRC Approval Will the proposed change determined to require prior NRC approval be either revised or rejected?

x D If No, then initiate a License Amendment Request in accordance 10 CFR 50.90 and AD-LS-ALL-0002, Regulatory Correspondence, and include the tracking number:

Part VIII: Signatures:

EP CFAM Final Approval Is required for changes affecting risk significant planning standard:

10 CFR . * *.* *

• r Name (Print): . . . . . . (] me (Print):

• Reilard o. Burm Date:** 11/,0 . If the proposed activity Is a change to the E-Plan. or implementing procedures, then create two EREG General: .. ** *

• Assignments;

• . . * * *. . . . . . . . . . .

• One for EP to provide the*10 CFR summary of the analysis;.or the completed 10 CFR 50.54(q);

• Ii* to Licensing;

• ** * * * * * * * * *One for licensing to submit the 10 CFR50.54(q) information to the NRC within 30 days after the change * .

• . *

• is put In effecL * * * -41

• QA RECORD-. .. . . . *-* *-**-* *-* -... . . . . . -: -* -* . . . . . . . . . .. t*".

ATTACHMENT 6 Page 1of4 << 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL) and EAL Bases Validation and Verification (V&V) Form >> Screening or Evi:lluation_Number::-':*

., ' ,, ./* :_**;"*-. , ', , ...... " :::. ' ' ..... ,.:**** " : .. "' : '.': (' ' Part I. Identification of ICs and EALs Affected by Proposed Change: Replaced BASES for EALs with the NRCapproved NEI 99-01rev06 Technical BASES Revision 1 Final in Section D of the Emergency Plan. 10CFRS0.54q performed per EREG #2069911.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:

A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

Changed "SIG tube RUPTURE" to "SIG tube leakage" Definition(s):

Deleted '"RUPTURE" This is NOT a change to an EAL, this is only a change to a BASES description.

---' --' -_, ' .* . ., .,. ',. " *. .. --:

.. ( '*'*,.*'*:':

• .',.***

• . ;'***. " : ., .. _ ... ,: .. , ,..:;:.,,.;

.. ;':* ., ,' *' ., ... ,*., ' ,. *.'.* " ... ' .. *'.: * .. * *;* .... Part II. Determination of Validation Method by Site EP Manager: In-Plant Walkdown D Tabletop D Training D Other (Specify)

Validation of LAR x Simulator D NA D EP Manager Name (Print):

Signature:

Date: Kevin L Murray .IA Jl-/o-/b

./ I**-.**., ,,, : . ,. r**;: ,. :'. .. :( *:::, .. :_:_,. ,, ".'*,,_.*:_:'.':* " **1.* ,_ *, :.: )* .. **:* *;' :*.

 : *,' ' * ** 0' *'-" ,* . ., -. *.-t' *;' Part Ill. Validation. (Answers marked No require resolution)

'" :. :-: :.

• Validation Oue.stion<.

• Yes :

NA, .. :. Resolution .ind Co(Tlments:*

,* r : . . ,, ' ... " . ._,' .-. Readouts, alarms, indications.

etc., available in D D x the Control Room? Monitor, gauge, etc., designations are correct? D D x Are correct units of measure displayed on the D D x monitor, oauQe, etc.? All values are within instrumentation display D D x range? Is instrument display finite enough to distinQuish b*etween values? D D x No miscellaneous issues were identified during v n n "' ..... ...... walkdown correct? ' *,:, ***:**' " ... *" .. :: l,,' "' Part IV. Verification (Answers marked No require resolution)

Validation Question Yes No NA " Resolution and Comments*

Is the IC/EAL change easy to use and does it flow well? Is sequencing logical and correct? D D Is it written to appropriate level of detail and unambiguous?

Is the IC/EAL Matrix legible and easy to use? D D Are correct units of measure displayed on the D D monitor, gauQe, etc.? Part IV. Verification (cont.) (No answers require resolution))

Instrumentation; Plant Computer System (PCS); and/or Plant Process Computer System (PPCS) points specified?

• Correct instrument?

• Correct units

• Adequate instrument range?

• Display unit readable?

• Proper significant digits?

• Instrument number and noun name provided?

• Consistent with operations procedures?

References specified in EAL Technical Basis current and updated and source documents for inputs have been identified and verified to be appropriate for use? Does the change avoid human performance challenges, latent weaknesses, and human performance traps?

• No vague or missing critical detail(s} .

• Decisions are not over-reliant on knowledge for successful performance Modifications, Emergency Plan, EAL Technical Basis, reference manual and procedure revisions, setpoint changes, software changes, training, etc. are appropriately scheduled to correspond to the EAL revision?

Are alarm setpoints equal to or below EAL thresholds?

Do radiation monitor setpoints account for backQround?

-'. .. .. . . ,* ' ,,,. " .. , ,-,, " . . .. Part V. Comments: D D x D -* x D x D D D D D , ,' ' x x x NA': X . D ------. -----D D x x . .. i' -_-, ; o**. *** :>< .. *' ,., ,. ., .. ,.--.\,". '.*

,* Part Vt Completion Review and Approval Signatures , Validation and. Verification (Print Names)(NOte1]:

' .

Signatures:._*.

Renaro O.Bwrls. _.*

Site EP Manager Review (Print Name): * . _Kewj L Mumy .* Qualified Emergency Coordinator (Print Name): .. Jtmmj*: Glenn older Signature

' ..

• Engineering Review t Name) {Note2],:

_* .* _ .*. -. ;V(-A *.*. *' .. . . . . . . . . . . . : . . . *Engineering Signa ure:

• PSA Review (Print;im

.. EP _CFAM Review: (P rnt Name): EP CFAM Signature:; . . M1 Jfftl -' / " --* -* -* ** c -.-*

• .li:tM [BNP; CR3; HNP, NP Final PNSC Approval (Print [BNP, CR3, NP 1 Rl\IP] PNSC Signalure Name) r-<i£4 * * * /'*I', 'J1

• Date: *

• lllioJtb . Date: . . . . t/;.}6-1&

.... Date:,.*'* . ll/6 16 . Date: .* *

• .. _oa.te:/ ... -*.* .n1/A: Date: .*.

" . ' Part VII. NRC Emergency Plan and Implementing Procedure Submittal Actions If the proposed activity is a ch.ange to the E-Plan or implementing procedures, then create two EREG General Assignments.

• One for EP to provide the 10 CFR 50.54(q) summary of the analysis, or the completed 10 CFR50.54(q), to Licensing.

• One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect. Notes: 1. Validation and Verification can be performed by same individual but must be:

• Qualified in the subject matter

• Separate from the author of change

• A cross-discipline reviewer 2. [BNP, CR3, HNP, RNP] System specific Engineering Review is required for EAL changes related to process equipment such as radiological instruments and environmental monitoring.

3. [BNP, HNP, RNP] PSA review is required for EAL changes to ensure any potential or actual impact to PSA calculations or assumptions are adequately addressed. (Not applicable to CR3) QA RECORD