Text

Site Emergency Plan
	ML18036A535
Person / Time
Site:	Mcguire, McGuire
Issue date:	01/26/2018
From:	; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
References
	MNS-18-009
	Download: ML18036A535 (359)
	v • d • e

DUKE ENERGY McGUIRE NUCLEAR SITE EMERGENCY PLAN SITE VICE DATE APPROVED:

REVISION 18-1: January, 2018 EFFECTIVE DATE: January, 2018 ORIGINAL DATE: August 25, 1980

DUKE ENERGY COMPANY McGUIRE NUCLEAR SITE EMERGENCY PLAN REVISION LIST August 25, 1980 Date Issued Revision 37, March 1992 Change 1, October, 1980 Revision 92-1, August 1992 Change 2, February, 1981 Revision 92-2, October 1992 Change 3, June, 1981 Rev. 93-1, April 1993 Change 4, August, 1981 Rev.93-2,June, 1993 Revision 1, November 16, 1981 Rev. 93-3, December 1993 Revision 2, February, 1982 Rev.94-1,January, 1994 Revision 3, february, 1982 Rev.94-2,June, 1994 Revision 4, April, 1982 Rev. 94-3, August 1994 Revision 5, June, 1982 Rev. 94-4, October 1994 Revision 6, July, 1982 Rev. 95-1, February 1995 Revision 7, September, 1982 Rev. 95-2, April 1995 Revision 8, November, 1982 Rev. 96-1, April 1996 Revision 9, January, 1983 Rev. 96-2, July 1996 Revision 10, February, 1983 Rev. 97-1, April 1997 Revision 11, June, 1983 Rev. 97-2, May 1997 Revision 12, November, 1983 Rev. 97-3, July, 1997 Revision 13, March, 1984 Rev.98-1,January, 1998 Revision 14, August, 1984 Rev.98-2,February, 1998 Revision 15, January, 1985 Rev.98-3,May, 1998 Revision 16, March, 1985 Rev. 98-4, July, 1998 Revision 17, May, 1985 Rev. 98-5, August, 1998 Revision 18, November, 1985 Rev. 98-6, November, 1998 Revision 19, January, 1986 Rev. 99-1, March, 1999 Revision 20, July, 1986 Rev. 99-2, July, 1999 Revision 21, May, 1987 Rev. 99-3 November, 1999 Revision 22, June, 1987 Rev. 00-1, April, 2000 Revision 23, November, 1987 Rev.00-2,May,2000 Revision 24, March, 1988 Rev. 00-3, November, 2000 Revision 25, July, 1988 Rev.Ol-1,January,2001 Revision 26, July, 1989 Rev.Ol-2,June,2001 Revision 27, September, 1989 Rev. 02-1, March, 2002 Revision 28, October, 1989 Rev. 02-2, August, 2002 Revision 29, November, 1989 Rev. 03-1, April, 2003 Revision 30, March, 1990 Rev.03-2,June,2003 Revision 31, April, 1991 Rev.04-l,February,2004 Revision 32, July, 1991 Rev. 04-2, July, 2004 Revision 33, September, 1991 Rev. 05-1, July, 2005 Revision 34, October, 1991 Rev.06-l,January,2006 Revision 35;December, 1991 Rev. 06-2, September, 2006 Revision 36, January, 1992 Rev.07-l,May,2007 1

Rev. 18-1 January, 2018

Rev. 07-2, December, 2007 Rev. 08-1, September, 2008 Rev. 09-1, July, 2009 Rev. 09-2, December, 2009 Rev. 10-1,I\1ay,2010 Rev. 10-2, November, 2010 Rev. 11-1, I\1arch, 2011 Rev. 11-2, August, 2011 Rev. 11-3, October, 2011 Rev. 12-l,I\1ay,2012 Rev. 12-2,June,2012 Rev. 12-3, November, 2012 Rev. 12-4, December, 2012 Rev. 13-l,I\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. 15-1,I\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 Rev. 17-1, September, 2017 Rev. 18-1,January,2018 2

Rev. 18-1 January, 2018

McGuire Emergency Plan List of Effective Pages Emergency Plan Approval Cover Sheet Coversheet Rev. 18-1 January,2018 Emergency Plan Revision List Page 1 thru 2 Rev. 18-1 January, 2018 I_

Table of Contents Page 1 thru 6 Rev. 18-1 January, 2018 l List of Figures Page 1 thru2 Rev. 18-1 January, 2018 1-.

Introduction Pages i-1 thru i-7 Rev. 13-3 October, 2013 A. Assignment of Responsibility Pages A-1 thru A-5 Rev. 16-2 April, 2016 B. Onsite Emergency Organization Pages B-1 thru B-15 Rev. 18-1 January, 2018 r

C. Emergency Response Support & Resources Pages C-1 thru C-2 Rev. 18-1 January, 2018 D. Emergency Class System/EAL Basis Document Pages D-1 thru D-259 Rev. 18-1 January, 2018 E. Notification Methods & Procedures Pages E-1 thru E-11 Rev. 16-3 September, 2016 Rev. 18-1 January, 2018

McGuire Emergency Plan List of Effective Pages F. Emergency Communications Pages F-1 thru F-7 Rev. 17-~ September, 2017 G. Public Education & Information Pages G-1 thru G-2 Rev. 18-1 January, 2018 H. Emergency Facility & Equipment Pages H-1 thru H-17 Rev. 18-1 January, 2018 I. Accident Assessment Pages I-1 thru I-3 Rev. 18-1 January, 2018 J. Protective Response Pages J-1 thru J-22 Rev. 18-1 January, 2018 K. Radiological Exposure Control Pages K-1 thru K-4 Rev. 18-1 January, 2018 L. Medical & Public Health Support Pages L-1 thru L-2 Rev. 18-1 January, 2018 M. Recovery & Re-entry Planning Pages M-1 thru M-5 Rev. 18-1 January, 2018 N. Exercises & Drills Pages N-1 thru N-3 Rev. 17-1 September, 2017 Rev. 18-1 January, 2018

McGuire Emergency Plan List of Effective Pages

0. Radiological Emergency Response Training Pages 0-1 thru 0-2 Rev. 17-1 September, 2017 P. Development Periodic Review & Distribution of Emergency Plans Pages P-1 thru P-10 Rev. 18-1 J anua.I)'., 2018 O. Appendices Index Pages Q-1 Rev. 17-1 September, 2017 Appendix 1 Def"mitions Rev. 18-1

January, 2018 Pages 1 thru 3 Appe1_1dix 2 Meteorological Program Rev. 15-3 October,2015 Pages 1 thru 4 Appendix 3 Alert & Notifications System Rev. 15-2 July, 2015 Pages 1 thru 5 Appendix 4 Evacuation Time Estimates Rev. 15-2 July, 2015 Page 1 Appendix 5 Agreement Letters Rev. 15-3 October, 2015 Pages 1 thru 4 Appendix 6 Emergency Plan Distribution Rev. 15-2 July, 2015 Pages 1 thru 4 Rev. 18-1 January, 2018

McGuire Emergency Plan List of Effective Pages Appendix 7 SPCC Plan Coversheet Rev. 15-1 May,2015 Table of Contents Rev. 15-1 May,2015 Pages 1 thru 75 Rev. 15-1 May,2015 Appendix 8 Hazardous Waste Contingency Plan Pages 1 thru 19 Rev. 13-1 March, 2013 Appendix 9 Hazardous Materials Response Plan Pages 1 thru 14 Rev. 17-1 September, *2017 Rev. 18-1 January, 2018

DUKE ENERGY MCGUIRE NUCLEAR SITE EMERGENCY PLAN Table of Contents Page#

1. Introduction A. Purpose i-1 B. Scope i-1 C. Planning Basis i-3 A. Assignment of Responsibility A.La Organization A-1 A.1.b Concept of Operations A-3 A.Le Block Diagram Interrelationships A-3 A.Ld Key Decisionmaking A-3 A.Le 24 Hour Emergency Response A-4 A.2.a Responsibility for and Functions of Emergency A-4 Response Org~nization A.2.b Legal Basis for Authority A-4 A.3 Agreement Letters for Emergency Response Support A-4 A.4 Individuals Responsible for Continuity of Resources A-5 B. On-site Emergency Organization B.1 Plant Staff Under Emergency Conditions B-1 B.2 Emergency Coordinator B-1 B.3 Emergency Coordinator (line of succession) B-1 B.4 Functional Responsibilities of Emergency Coordinator B-2 B.5 Minimum Staffing Requirements B-2 B.6 On-site Functional Area Interfaces B-2 B.7 Augmented Support of On-site Emergency Organization B-2 B.8 Contractor and Private Organizations B-3 thru B-4 B.9 Local Agency Support Services B-4 B.9.a Law Enforcement, Emergency Traffic Control, Related B-4 Police Matters B.9.b Early Warning or Evacuation of the Populace B-5 B.9.c Radiological Emergency Monitoring Assistance B-5 B.9.d Hospitals, Medical Support B-5 B.9.e Ambulance Service B-5 B.9.f Fire-fighting B-5 B.9.g Public Health and Safety, Evaluation of the B-5 Radiological Situation B.9.h Local, State and Federal Support Responsibilities B-6 Rev. 18-1 September, 2017 1

Table of Contents Page#

C. Emergency Response Support and Resources C.l.a Individuals Authorized to Request Federal Assistance C-1 C.1.b Federal Resources Arrival Time C-1 C.l.c Emergency Operations Facility Resources C-1 Available to Federal Response Organizations C.2.a State and County Representation at the Emergency C-1 Operations Facility (EOF)

C.2.b Licensee Representation at the Off-Site EOC's C-1 C.3 Radiological Laboratories-Availability and Capability C-2 C.4 Emergency Support From Other Organizations C-2 D. Emergency Classification System/EAL Basis Document Emergency Classification System/EAL D-1 thru D-262 Basis Document E. Notification Methodology E.l Notification of Response Organization E-1 E.2 Activation of Emergency Organization E-1 E.2.a Notification of Unusual Event E-1 E.2.b Alert E-2 E.2.c Site Area Emergency E-4 E.2.d General Emergency E-6 E.3 Emergency Message Format (Initial) E-8 E.4 Emergency Message Format (Follow-up) E-9 E.5 State and Local Organizations-Disseminating Public E-9 Information E.6 Alert and Notification System E-9 E.7 Supporting Information for Public Information Message E-9 F. Emergency Communications F.l.a 24 Hour Notification Capability F-1 F.l.b Communications With State/Local Governments F-1 F.l.c Communications With Federal Organizations F-2 F.l.d Communications Between Site, EOF, EOC's F-2 and Monitoring Team F.l.e Activation of Emergency Personnel F-2 F.l.f Communications Between NRC, EOC and F-2 Monitoring Teams F.1.g ERDS Data Transfer F-2 F.2 Medical Support Communications F-3 F.3 Communications System Testing F-3 Rev. 18-1 January, 2018 2

Table of Contents *Page#

G. Public Education and Information G.1/G.2 Public Education and Information Program G-1 G.3.a News Group - Location and Contacts G-1 G.3.b News Group - Media Center G-1 G.4.a Public Spokesperson G-2 G.4.b Spokesperson Information Exchange G-2 G.4.c Rumor Control G-2 G.5 News Media Training Sessions G-2 H. Emergency Facilities and Equipment H.1 Technical Support Center/Operations Support Center H-1 H.1.a Control Room H-1 H.l.b Technical Support Center (TSC) H-1 H.l.c Operations Support Center (OSC) H-2 H.1.d Alternate Facilities H-2 H.2 Emergency Operations Facility (EOF) H-2 thruH-3 H.3 State and Local Government Emergency Operations Center H-4 H.4 Activation and Staffing H-4 H.5 Assessment Actions H-4 H.5.a Meteorological, Hydrologic and Seismic H-4 thru H-5 H.5.b Radiological Monitors H-6 H.5.c Plant Parameters H-6 H.5.d Fire Detection H-6 H.6 Data, Monitoring Equipment and Analysis Facilities H-6 H.7 Off-site Radiological Monitoring H-7 H.8 Meteorology Instrumentation and Procedures H-7 H.9 Operations Support Center H-7 H.10 Emergency Equipment/Instrumentation Inspection, H-7 Inventory, Operational Check, Calibration H.11 Emergency Kits H-7 H.12 Receipt and Analysis of Field Monitoring Data H-7 I. Accident Assessment 1.1 Emergency Action Level Procedures 1-1 1.2 On-site Capability and Resources to Provide Initial 1-1 Values and Continuing Assessment 1.2.a Post Accident Sampling 1-1 1.2.b Radiation and Effluent Monitors 1-1 1.2.c In-plant Iodine Instrumentation 1-2 1.3.a/ Method for Determining Release Source Term 1-2 1.3.b 1.4 Effluent Monitor Readings Vs On-site/Off-site Exposure 1-2 1.5 Meteorological Information Availability 1-2 1.6 Release Rates/Projected Doses for Offscale Instrumentation 1-2 Rev. 18-1 January, 2018 3

Table of Contents Page#

1.7/ Field Monitoring Within EPZ 1-3 1.8 1.9 Detect and Measure Radioiodine Concentration in the EPZ 1-3 1.10 Relationship Between Contamination Levels 1-3 and Integrated Dose/Dose Rates 1.11 Plume Tracking 1-3 J. Protective Response J.l.a Onsite Alerting and Notification J-1 thru J.l.d J.2 Evacuation Routes and Transportation J-1 J.3 Personnel Monitoring J-1 J.4 Site Evacuation Procedures - Decontamination J-2 J.5 Personnel Accountability J-2 J.6 Protective Equipment Breathing Apparatus, Protective J-2 Clothing, KI J.7 Protective Action Recommendations J-3 J.8 Evacuation Time Estimates J-4 J.9 Implementing Protective Measures J-4 J.10 Implementation of Protective Measures for Plume J-5 Exposure Pathway J.10.a EPZ Maps J-6 J.10.b EPZ Population Distribution Map J-6 J.10.c EPZ Population Alerting and Notification J-6 J.10.d EPZ Protecting Immobile Persons J-6 J.10.e Use ofRadioprotective Drugs for Persons in EPZ J-6 J.10.f Conditions for Use ofRadioprotective Drugs J-6 J.10.g State/County Relocation Plans J-6 J.10.h Relocation Center Locations J-6 J.10.i Evacuation Route - Traffic Capacities J-6 J.10.j Evacuated Area Access Control J-6 J.10.k Planning for Contingencies in Evacuation J-6 J.10.1 State/County Evacuation Time Estimates J-6 J.10.m Bases for Protective Action Recommendations J-6 J.11 Ingestion Pathway Planning J-7 J.12 Relocation Center - Registering and Monitoring J-7 Rev. 18-1 January, 2018 4

Table of Contents Page#

K. Radiological Exposure Control K.1 Onsite Exposure Guidelines K-1 K.2 Doses in Excess of 10CFR Part 20 K-1 K.3 Emergency Personnel Exposure and Records K-1 K.3.a Distribution of Dosimetry K-1 K.3.b Dose Records K-2 K.4 State/Local Plan for Authorizing Doses Exceeding PA G's K-2 K.5 Decontamination K-2 K.5.a Action Levels for Determining the Need for Decontamination K-2 K.5.b Radiological Decontamination K-2 K.6 Contamination Control Measures K-2 K.6.a Area Access Control K-2 K.6.b Drinking Water and Food Supplies K-2 K.6.c Recovery Efforts K-3 K.7 Decontamination of Personnel at Relocation K-3 Assembly Area L. Medical and Public Health Support L.1 Hospital and Medical Support L-1 L.2 On-site First Aid Capability L-1 L.3 Public, Private, Military Hospitals, Emergency Medical L-1 Facilities L.4 Transport of Accident Victims L-2 M. Recovery and Reentry Planning and Post-Accident Operations M.1 Recovery/Reentry Plans and Procedures M-1 M.1.a Outline of Site Recovery Plans M-1 M.1.b Outline of Recovery Plans M-2 M.2 Recovery Organization M-3 M.3 Information to Members of Recovery Organization M-4 M.4 Total Population Exposure Estimates M-4 N. Exercises and Drills N.1.a Exercises N-1 N.1.b Exercise Scenario/Response N-1 N.2 Drills N-1 N.2.a Communications N-1 N.2.b Fire Drills N-2 N.2.c Medical Emergency Drills N-2 N.2.d Radiological Monitoring Drills N-2 N.2.e Radiation Protection Drills N-2 N.3 Exercise and Drill Execution N-3 N.4 Exercise Critique N-3 N.5 Critique Action Items N-3 Rev. 18-1 January, 2018 5

Table of Contents Page#

0. Radiological Emergency Response Training 0.1 Offsite Agency Training 0-1 0.1.a Emergency Response Training (Offsite Agency) 0-1 0.1.b Off-site Support Agency - Participation in Training 0-1 0.2 Site Organization Training 0-1 0.3 First Aid Training 0-1 0.4 Training For Radiological Emergency Response 0-2 Personnel 0.5 Training Period 0-2 P. Responsibility for the Planning Effort P.l Emergency Planning Staff Training P-1 P.2 Emergency Response Planning P-1 P.3 Site Emergency Planning Manager P-1 P.4 Review of Emergency Plan P-1 P.5 Distribution of Revised Plans P-1 P.6 Supporting Plans P-2 P.7 Implementing Procedures P-2 P.8 Table of Contents P-2 P.9 Audit of Emergency Plan P-2 P.10 Telephone Number Updates P-3 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 Rev. 18-1 January, 2018 6

LIST OF FIGURES FIGURE NO. TITLE PAGE#

i-1 10 MileEPZ i-6 i-2 50 MileEPZ i-7 A-1 Responsibility for Emergency Response Functions A-6 B-la Minimum On-Shift ERO Staffing Requirements for Emergencies B-7 B-lb Minimum Augmented ERO Staffing Requirements for Emergencies B-8 &9 B-2 Site Emergency Organization B-10 B-3 OSC Site Emergency Organization B-11 B-4 EOF Site Emergency Organization B-12 B-5 Joint Information Center (JIC B-13 B-6 Inter-Relationships of Response Organizations - B-14 Unusual Event B-7 Inter-Relationships of Response Organizations - B-15 Alert, Site Area Emergency, General Emergency, General Emergency E-1 Emergency Notification E-10 F-1 Emergency Communication Layout Prior to TSC/EOF Activation F-4 F-2 Emergency Communication Layout After TSC Activation, F-5 and During EOF Activation F-3 Emergency Response Radio System F-6 F-4 Emergency Operations Facility Communications F-7 G-1 Corporate Communications Initial Emergency Response Organization G-3 G-2 Corporate Communications Emergency Response Organization G-4 H-1 McGuire Nuclear Site Technical Support Center H-8 H-2 McGuire Nuclear Site Operations Support Center H-9 H-3 McGuire/Catawba/Oconee EOF General Location H-10 Rev. 18-1 1 January, 2018

LIST OF FIGURES FIGURE NO. TITLE PAGE#

H-4 McGuire/Catawba/Oconee EOF Location H-11 H-5 McGuire/BOP General Arrangement H-12 H-6 Media Center H-13 H-7 Joint Information Center H-14 H-8 Generalized Met System H-15 H-9 Alternate TSC H-16 H-10 Alternate OSC H-17 J-1 Guidance for Offsite Protective Actions J-8 J-2 Des_cription of Evacuation Regions J-11 J-3 MNS ETE Based on 2010 Census - 90% Effected Population J-13 J-4 MNS ETE Based on 2010 Census - 100% Effected Population J-17 J-5 Evacuation Route Map for MNS J-21 J-6 Summary of Population and Demand J-22 K-1 Emergency Workers Exposure Guidelines K-4 M-1 Recovery Organization M-5 P-1 Supporting Plans P-4 P-2 Emergency Plan Implementing Procedures P-5 thru P-8 P-3 Emergency Plan Implementing Procedures Distribution P-9 P-10 2-1 Meteorological Parameters of the Upgraded System Q-9 Q-1 Siren Range in Feet Q-13 Q-2 Siren Locations Q-14 Rev. 18-1 2 January, 2018

B. On-Site Emergency Organization B.l Plant Staff Under Emergency Conditions Figures B B-5 shows the emergency organization of plant staff personnel for all shifts.

The relationship of these personnel to their normal responsibilities and duties is unchanged during an emergency condition.

B.2 Emergency Coordinator Initial activities at McGuire during any emergency condition are directed by the Shift Manager from the Control Room. The Shift Manager shall assume the functions of the Emergency Coordinator until the arrival of the Station Manager/Designee at which time the Station Manager/Designee will assume the functions of the Emergency Coordinator. The Emergency Coordinator will have the authority and the responsibility to immediately and unilaterally initiate any emergency actions including:

a. Provide protective action recommendations to authorities responsible for implementing off-site emergency measures, implement event classification, notification, and event escalation/de-escalation/termination. THIS AUTHORITY SHALL NOT BE DELEGATED TO OTHER ELEMENTS OF THE EMERGENCY ORGANIZATION EXCEPT FOR EVENT CLASSIFICATIONS WHICH ARE PERFORMED AT THE STATION BY AN EMERGENCY COORDINATOR WHEN EOF IS ACTIVATED.

b. Notification and activation of the Site, Corporate, County/City, North Carolina and the Nuclear Regulatory Commission emergency organizations having a response role.

c. Continued assessment of actual or potential consequences both on-site and off-site throughout the evolution of the emergency condition.

d. Effective implementation of emergency measures in the environs including protective actions for affected areas, implementation of emergency monitoring teams and facilities to evaluate the environmental consequences of the emergency condition, prompt notification and communications with off-site authorities.

e. Continued maintenance of an adequate state of emergency preparedness until the emergency situation has been effectively managed and the site is returned to a normal or safe operating condition.

B.3 Emergency Coordinator (Line of Succession)

The Emergency Coordinator function as described above in paragraph B.2 will later be assumed by the TSC Emergency Coordinator and/or EOF Director at the Emergency Operations Facility as this organization is staffed and ready to take over its functions.

This assumption of the Emergency Coordinator functions will take place for the Alert, Site Area Emergency and General Emergency categories.

Rev. 18-1 B-1 January, 2018

IF AT ANY TIME the EOF is Activated, THEN the following applies:

Classification of events are performed by either the TSC or Control Room

Immediate communication to the EOF is required upon upgrade of a classification of an event by either the TSC or Control Room

Notifications to Offsite Agencies are performed by the EOF

Protective Action Recommendations (PAR) are performed by the EOF B.4 Functional Responsibilities of the Emergency Coordinator The functional responsibilities of the Emergency Coordinator are described in paragraph B.2. Protective Action recommendations to state and local authorities is initially vested with the Shift Manager/ Emergency Coordinator. As the Emergency Operations Facility (EOF) becomes activated, the EOF Director is the person who is responsible for making protective action recommendations.

B.5 Minimum Staffing Requirements The positions, title and major tasks to be performed by the persons assigned to the functional areas of emergency activity at the site are described in the Emergency Planning Group Manual, Section 1.1. These assignments shall cover the emergency functions in Figure B-l (alb). The minimum on-shift staffing reflective of 2 Units in operation is as indicated in Figure B-1 a. The capability to augment on-shift resources after declaration of an emergency is as indicated in Figure B-1 b. The functional tasks to be performed by persons assigned to the areas of emergency* activity are as designated in Emergency Planning Group Manual, Section 1.1.

A detailed analysis demonstrating that on-shift personnel assigned emergency plan implementation functions are not assigned responsibilities that would prevent the timely*

performance of their assigned functions as specified in Figure B-1.a is located in MNS-OSSA-12212012 Rev: 0. .

B.6 On-site Functional Area Interfaces Figures B B-7 describe and specify the interfaces between and among the functional areas of emergency activity, licensee headquarters support, local services support, and state/local government response organizations. Figure B-6 is for use prior to activation of the EOF. Figure B-7 is for use after the EOF is established.

B.7 Augmented Support of On-site Emergency Organization Upon declaration of an Alert, Site Area Emergency or General Emergency, the EOF organization will be alerted and personnel will report to the EOF as soon as possible. The EOF organization is described in Emergency Planning Implementation Procedures. The Communications organization is described in Section G.3.a. Figure B-4 shows the minimum staff required to declare the EOF operational. The EOF will be staffed using 75 minutes as a goal for the minimum staff to be in place and operational.

Rev. 18-1 B-2 January, 2018

In addition to the minimum staff shown in Figure B-4, other EOF personnel shown are expected to report to the EOF to augment the minimum staff. This augmentation would occur gradually and would range from a few minutes to a few hours depending on the proximity of the personnel to the EOF.

The organization identified in this section is capable of continuous (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) operations for a protracted period. The individual responsible for assuring continuity of resources is the EOF Director. Each group's operational plan is specified in the Emergency Plan.

B.8 Contractor and Private Organizations The Institute of Nuclear Power Operations (INPO) serves as* a clearinghouse for industry wide support during an emergency. When notified of an emergency situation at a nuclear plant, INPO will provide emergency response as requested. Contact will be made with INPO through the EOF. INPO will be able to provide the following emergency support functions:

a. Assistance to the affected utility in locating sources of emergency manpower and equipment.

b. Analysis of the operational aspects of the incident.

c. Dissemination to member utilities of information concerning the incident.

d. . Organization of industry experts who could advise on technical matters.

If requested, one or more suitably qualified members of the INPO staff will report to the EOF Director and will assist in coordinating INPO's response to the emergency.

The State of North Carolina The response by the State of North Carolina to an emergency development is described in the North Carolina Radiological Emergency Response Plan in Support of McGuire Nuclear Site and in their plan for Catawba Nuclear Site.

The principal state agency for mobilization of State resources to cope with an emergency is the Division of Emergency Management. This agency is supported by the Division of Radiation Protection for radiological assessment and protection functions, and by other State agencies.

The state organization, when it is mobilized as the State Emergency Response Team (SERT), becomes the primary response authority. For an emergency at McGuire, the SERT organization is established in the Emergency Operations Center in Raleigh, N.C.

Nuclear Regulatory Commission The response provided by the NRC to an emergency developing at a Duke nuclear site is described in the NRC Region II Emergency Plan. The representative of the NRC who would provide input to the EOF Director is the Director of Region II. He is provided work space and a telephone in the EOF.

Rev. 18-1 B-3 January, 2018

The role of the NRC in an emergency situation is to provide oversight and recommendations on licensee actions.

County Governments

. In an emergency situation at a nuclear site, county governments are immediately notified of the accident. They have the primary responsibility for the protection of the citizens within the county boundaries. The principal Duke Energy contact with county government is through the Emergency Management Director or designee. This contact will be maintained by the TSC until relieved by EOF Offsite Agency Communicator.

It is recognized that the county council, the chief executive of the county, and mayors of local communities have responsibilities in an emergency situation as well. The Government Liaison on the staff of the Joint Information Center serves as the primary Duke Energy contact with these people.

American Nuclear Insurers (AND ANI will be notified of emergency conditions by the EOF Services. ANl's response group would set up claims payments and other such capabilities at facilities appropriate to the emergency.

Contractor The contractor who may be requested to respond is Westinghouse. Westinghouse will operate from Pittsburgh, Pennsylvania, with a small contingent at the plant.

B.9 Local Agency Support Service State, local and county agencies responsible for public health and safety work through the Emergency Management Agency Emergency Operations Center in the affected county until the State Emergency Response Team assumes control. The EOF coordinates with the agencies necessary to support the emergency condition. Agencies that have agreed to provide support, as necessary to McGuire Nuclear Site and surrounding areas, are listed below:

B.9 .a Law Enforcement, Emergency Traffic Control, Related Police Matters

1. Charlotte-Mecklenburg Police

2. North Carolina Highway Patrol B.9.b Early Warning or Evacuation of the Populace

1. Department of Emergency Management, Catawba County, (Newton, NC)

2. Department of Emergency Management, Gaston County (Gastonia, NC)

3. Department of Emergency Management, Mecklenburg County (Charlotte, NC)

4. Civil Preparedness Agency of Iredell County (Statesville, NC)

Rev. 18-1 B-4 January, 2018

5. Department of Emergency Management, Cabarrus County (Concord, NC)

6. Department of Emergency Management, Lincoln County (Lincolnton NC)

7. North Carolina Department of Public Safety (Raleigh, NC)

B.9.c Radiological Emergency Monitoring Assistance

1. US/DOE Radiological Assistance Team, Savannah River Operations Office (Aiken, SC)

2. North Carolina Department of Environmental Natural Resources, Division of Radiation Protection (Raleigh, NC)

3. Civil Air Patrol, North Carolina Wing (Charlotte, NC)

B.9.d Hospitals, Medical Support

1. Carolinas Medical Center (Charlotte, NC)

2. REACTS Facility, DOE (Oak Ridge, TN)

B.9.e Ambulance Service

1. MEDIC (Cornelius, NC)

B.9.f Fire-Fighting

1. Cornelius Volunteer Fire Department (Cornelius, NC)

2. Huntersville Fire Department (Huntersville, NC)

3. Mecklenburg County Fire Marshall B.9.g Public Health and Safety, Evaluation of the Radiological Situation.

1. North Carolina Department of Environmental Natural Resources, Division of Radiation Protection (Raleigh, NC)

Rev. 18-1 B-5 January, 2018

B.9.h Local, State and Federal Support Responsibilities Agreements have been made with local, state and federal agencies to provide fire protection, medical support, ambulance service, rescue service and hostile action response. Implementation of the emergency plans of the Emergency Management Agencies of six adjacent counties will provide assistance and logistics support if evacuation of portions of the ten mile EPZ becomes necessary. The emergency plans of the Emergency Management Agencies in Mecklenburg County where the site is located, and in Gaston, Catawba, Iredell, Lincoln and Cabarrus Counties, North Carolina, as they relate to the protection of the public who may be affected by an emergency at McGuire, all address the following aspects:

1. Notification of their own personnel and other agencies involved, including the Sherift's Department, the Highway Patrol, police, rescue squads, fire departments and the Red Cross.

2. Law enforcement and traffic control.

3. Notification or warning of persons in affected areas

4. Evacuation, as necessary, to designated schools or other public buildings out of the affected area, where shelter, food, overnight accommodations, communications, medical care; etc. would be made available.

5. Assistance and cooperation with related agencies in other counties, Duke Energy, and other state and federal agencies.

NOTE: Summary written agreements with the agencies that have various responsibilities for emergency preparedness support and for emergency response in the public domain are included in the Appendix 5.

Rev. 18-1 B-6 January, 2018

FIGUREB-la MCGUIRE NUCLEAR SITE MINIMUM ON-SHIFT ERO STAFFING REQUIREMENTS FOR EMERGENCIES CR Supervisor (SRO) 1

1. Plant Operations and Control Room Operator (RO) 4 Assessment of Non-Licensed Operator (NLO) 3 Operational Aspects (a)

WCCSRO 1

2. Emergency Direction and Command and Control Ops Shift Manager (SRO)

Control

3. Notification & !A.C..~1.1.:~.~~................ ................ .. . . . . . . Operator. (SRO/RO/NLO). .. . 1(br**

Communication -~~d=~?-!.~!~-***----*********--*-*--- g~:~:~~~.i~~g~g~~g~ ... 1

- - - - 1 (b) -

Dose As.~~~s~~.1.1.:!...................... .....................B:1.1....Q~~.1.i.Q~d In~!.Y.!.~~al ................................... 1

4. Radiological Assessment In-plant Surveys ..................................--*---* . RP. .Qualified . Individual-*-********* .......................-.. . . !... __ _

Onsite Surveys RP Qualified Individual -*---** 1 Chemistry Chemistry Technician 1 Tech Support - OPs Shift Technical Advisor 1

5. Plant System Engineering, Repair, and - Core Damage *shlti°T~.~~~~(Ady~so~--*-----:=::~1 (b)*:::~::

Repair Corrective Actions Mechanical Maintenance 1 Corrective Actions IAE Maintenance 2 Radiation Protection (such as access control, job

6. In-Plant P As RP Qualified Individual coverage and personnel monitoring)

Fire Brigade Lead (RO/SRO)

7. Fire Fighting (c) Fire Brigade Member (NLO)

Fire Brigade Member (SPOC)

8. 1st Aid and Rescue MERT (d)

9. Site Access Control and Security & Accountability SAS Operator Accountability Security Personnel (a) The Control Room staff complement is reflective of2 Units in operation.

(b) May be performed by an individual filling another position provided they are qualified to do the collateral function.

(c) The Fire Brigade requirement of five members is met by using four personnel from Operations (including the Fire Brigade Leader) and one person from SPOC (SLC 16.13.1).

(d) The Medical Emergency Response Team (MERT) is filled by Security Officers.

(e) Per Duke Energy MNS Security Plan.

Rev. 18-1 B-7 January, 2018

FIGUREB-lb MCGUIRE NUCLEAR SITE MINIMUM AUGMENTED ERO STAFFING REQUIREMENTS FOR EMERGENCIES Major Functional Area Major Task Position, Title or Expertise Capability for Additions****

45Min. 75Min.

Emergency Direction and TSC Emergency Coordinator 1 Control (Emergency Coordinator)**

Notification/Communication Notify Company Personnel, Off-site Agency Communicator 2 State, County, Federal Agencies and Maintain Communication Emergency Operations EOF Director Senior Manager 1 Facility (EOF) Radiological Dose Assessment Radiological Assessment 1 Accident Assessment and Manager Support Plant Status Accident Assessment Manager 1***

Access Control Electronic Card Reader #

Communications Off-site Agency Communicators 2 Off-site Surveys FMT Members (2 Teams) 4*****

Radiological Support and RP Coverage for Repair/ RP Qualified Individuals 6 Protective Actions Corrective Actions, Access Control, Search & Rescue, Radiochemistry, Contaminated Injury Medical Response, Personnel Monitoring, Dosimetry, Firefighting Out of Plant Surveys 1 1 In-Plant Surveys 1 1 Dose Assessment TSC Off-site Dose Assessor 1 Chem/Radwaste Radwaste Operator 1 Operations Plant System Engineering, Technical Support Core/Thermal Hydraulics 1***

Repair and Corrective Actions Electrical 1 Mechanical 1 Mechanical Maint. Tech. 1 Repair and Corrective IAE Technician 2 Actions Firefighting Fire Brigade Local Support Rescue Operations and First MERT Local Aid Support Rev. 18-1 B-8 January, 2018

FIGUREB-lb MCGUIRE NUCLEAR SITE MINIMUM AUGMENTED ERO STAFFING REQUIREMENTS FOR EMERGENCIES

- Management of the off-site Emergency Response will be assumed by the EOF Director when the Emergency Operations Facility is activated. Management of the on-site Emergency Response is assumed by the Station Manager/alternate acting as the Emergency Coordinator when the Technical Support Center and Operations Support Center are

activated.

- - The TSC Reactor Engineer and the Accident Assessment Manager in the EOF will provide additional support in the area of core/thermal hydraulics within 75 minutes.

- - - Consideration is given to the fact that most McGuire Site staff and support personnel do not choose or are unable to live within a radius of the site which will allow a response time of 30 minutes or less under ideal conditions. Factors such as weather conditions, road capacity and traffic density, limited housing (near site) and the distance to travel from residence to plant site indicate a realistic response time of from a few minutes to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 15 minutes for most employees. Consideration is also given to personnel on shift who are qualified and sufficient in number to handle any emergency condition until response personnel begin to arrive onsite in from a few minutes to one (1) hour and 15 minutes.

- - - - The Field Monitoring Teams will initially report to the Operations Support Center (OSC).

If needed, the Field Monitoring Teams will be dispatched from the Operations Support Center (OSC). Once the Emergency Operations Facility (EOF) Field Monitoring Coordinator is ready he/she will assume control of the Field Monitoring Teams. An FMT consists of one RP qualified individual and one vehicle driver.

An electronic card reader in conjunction with a posted building security officer fulfills the function for controlling access to the EOF during emergencies.

Rev. 18-1 B-9 January, 2018

FIGUREB-2 MCGUIRE NUCLEAR SITE SITE EMERGENCY ORGANIZATION (TSC)

, Emergency C<>ordinatof*

iii~"::t~ii,~r1~,~,i~c *** ,~.,.i,i:,;*

!1

'. .::'i>e<'.RPC: ~: ,*: ...*"

. ':.*,M~n,~~r1:t1:'**.

.* *. S~c1..1rity:;'* ...

.coorctinato~;.,

_,,'",:, ,,<,\, < ,' '-:.,-.>

,,,:,J.:t._.'.:./ '*'

- Bold Boxes indicates minimum staff position Rev. 18-1 B-10 January, 2018

FIGUREB-3 MCGUIRE NUCLEAR SITE SITE EMERGENCY ORGANIZATION (OSC}

. , OSC ~anage,r S'. 'S ' *s 75.min

?,.p,JL.,.,)\0.\':v_'/<, ,>t>l,:'.<-}f, * ;<'<

(',,,,JAE T~ch' (2):, ,

. .**\ !~!l!J~ii1:;,f;,;,

FMT Nlernber (:4)

.:,~~~fr~;~JitW.\P>x t;;~;~;

. 'AO (3) i' .* RPJech.,

Confrol Room :, .

-. ~,.'.fy;',;v 1L"<)i/t.**'r;*

~(,~!,<<f'

- Bold Boxes indicates minimum staff position Rev. 18-1 B-11 January, 2018

FIGUREB-4 MCGillRE NUCLEAR SITE EMERGENCY ORGANIZATION (EOF)

AccidentAssmt/

.,., *:, . Ma'nag~r/ ~':. ,

,;~; .(?r:,J~{~tJJ}!q '1~ttr.~f,:~ ,

Services,.Admin z~ *R.~t~J~~~~~~c>~<<2).

1 C *:,

))C>,

..<: it~f;~~~i'~~a;;t

- Bold Boxes indicates minimum staff position Rev. 18-1 B-12 January, 2018

FIGUREB-5 MCGUIRE NUCLEAR SITE JOINT INFORMATION CENTER (JIC)

Bold Box - Corporate Support Positions Ohers - Site Emergency Plan Positions Medi.a C.enter . lrifo .Gathering

..*, q?9J:tjio~t,~:~:r5 Jtjt.*:4~'(i!:Ji~~fc .* .

Rev. 18-1 B-13 January, 2018

FIGUREB-6 MCGUIRE NUCLEAR SITE INTER-RELATIONSHIPS OF RESPONSE ORGANIZATIONS UNUSUAL EVENT*

Shift Manager Shift Manager Control Room Determines Emergency Action Level Nuclear Operations Manager Plant Manager Local County NCEOC Nuclear Governments (Raleigh) Regulatory Commission

Does not require activation of any Emergency Response Organization Rev. 18-1 B-14 January, 2018

FIGUREB-7 MCGUIRE NUCLEAR SITE INTER-RELATIONSHIPS OF RESPONSE ORGANIZATIONS ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Operations Support Center Technical Support Center Control Room -

ON-SITE OFF-SITE I

Emergency Joint Operations Information INPO

,- Facility Center I

I State Federal

- Government (NC)

Government Agencies I I Local County Nuclear Regulatory FEMA Governments Commission DOE I

NSSS Supplier Local Emergency I I Response Agencies Rev. 18-1 B-15 January, 2018

C. Emergency Response Support and Resources C.l.a. Individuals Authorized to Request Federal Assistance Environmental radiological measurements are made in the field by radiological survey teams.

This information is used by the Radiological Assessment Manager to confirm environmental projections of doses and dose rates. If necessary to relieve Duke personnel, environmental surveillance support personnel from the DOE Radiological Assistance Plan may be requested by the Radiological Assessment Manager or the EOF Director.

C.l.b. Federal Resources -Arrival Time The Agreement letter between Duke Energy and DOE - Savannah River is found in Appendix 5. DOE emergency radiological assistance is expected within 3 to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from a call for these services at McGuire. (Driving and set up time - does not consider use of helicopter or other aerial means of transport).

NRC's full team from Region 2 would be on-site within 7-8 hours from declaration of an emergency at McGuire. Some portions of their team could arrive on-site much earlier by the use of helicopter transport from Atlanta.

C.1.c. Emergency Operations Facility Resources Available to Federal Response Organizations .

The following Duke Energy resources are available to support Federal emergency response from DOE - Savannah River.

McGuire Nuclear Site Airfield - Charlotte/Douglas International airport ( 40 minutes from site)

C.2.a State and County Representation at the Emergency Operations Facility (EOF)

The State and counties in the EPZ around Duke Energy nuclear facilities are provided for as described in Section H.

C.2.b Licensee Representation at the Off-Site EOC's Provisions have been made to dispatch representatives to principal off-site governmental Emergency Operations Centers (EOC's). The representatives act as liaisons to clarify the information contained in emergency notifications and provide an additional on-site link to the nuclear facility emergency response staffs.

Rev. 18-1 C-1 January, 2018

C.3 Radiological Laboratories - Availability and Capability Laboratory facilities include mobile emergency monitoring capabilities available through the NC Department of Environmental Health and Natural Resources, Division of Radiation Protection; and the DOE Radiological Assistance Team. In addition, the site has 2 emergency vehicles set up to provide monitoring capability. Fixed facilities are available for gross counting and spectral analysis in the site counting laboratory and at the nearby Duke Energy Applied Sciences Center (1/2 mile). Other facilities within the Duke System at .

Catawba Nuclear Site (45 miles) and at Oconee Nuclear Site (160 miles) could provide further analysis support within a short period of time (1-4 hours). The above radiological laboratories are available on a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day basis and could provide their services and equipment on demand.

C.4 Emergency Support From Other Organizations Other support can be provided by:

INPO Fixed Nuclear Facility, Voluntary Assistance Agreement Signatories DOE Savannah River Area Hospitals (see section B.9)

Volunteer Fire Departments (see section B.9)

Radiation Emergency Assistance Center/Training Site (REAC/TS)

Rev. 18-1 C-2 January, 2018

MCGUIRE NUCLEAR STATION EMERGENCY ACTION LEVEL TECHNICAL BASES D-1 Revision 18-1 January, 2018

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ...................................................................................................................................... 3 2.0 DISCUSSION ................................................................................................................................. 3 2.1 Background ............................................................................................................................. 3 2.2 Fission Product Barriers .......................................................................................................... 4 2.3 Fission Product Barrier Classification Criteria .......................................................................... 4 2.4 EAL Organization .................................................................................................................... 5 2.5 Technical Bases Information ................................................................................................... 7 2.6 Operating Mode Applicability ................................................................................................... 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ....................................................... 9 3.1 General Considerations ........................................................................................................... 9 3.2 Classification Methodology .................................................................................................... 10

4.0 REFERENCES

............................................................................................................................. 14 4.1 Developmental ...................................................................................................................... 14 4.2 Implementing ......................................................................................................................... 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ....................................................................... 15 6.0 MNS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ............................................................... 23 7.0 ATTACHMENTS ........................................................................................................................... 27 1 Emergency Action Level Technical Bases ................................................................... 28 Category R Abnormal Rad Release / Rad Effluent ............................................. 28 Category C Cold Shutdown / Refueling System Malfunction .............................. 71 Category H Hazards ........................................................................................... 108 Category S System Malfunction ......................................................................... 153 Category E ISFSI ............................................................................................... 194 Category F Fission Product Barrier Degradation ................................................ 198 2 Fission Product Barrier Loss/ Potential Loss Matrix and Bases ........................................................................................................200 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases ....................................... 254 D-2 Revision 18-1 January, 2018

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/EOF 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 decision-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 ML12326A805} (ref.4.1.1 ), MNS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

D-3 Revision 18-1 January, 2018

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:

Alert:

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 18-1 January, 2018

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 m!_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 EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identificati0h 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 18-1 January, 2018

EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory R - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 -Seismic Event 3 - Natural or Technological Hazard 4- Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Coordinator J,udgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI)

S - System Malfunction 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 F - Fission Product Barrier Degradation None C - Cold Shutdown / Refueling System 1 -NCS Level Malfunction 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 18-1 January, 2018

2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, 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:

Category Letter & Title Subcategory Number & Title Initiating Condition (IC)

Site-specific description of the generic IC given in NEI 99-01 Rev. 6.

EAL Identifier (enclosed in rectangle)

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

1. First character (letter): Corresponds to the EAL category as described above (R, C, 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 18-1 January, 2018

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 ~0.99 and reactor thermal power > 5%

2 Startup Keff ~0.99 and reactor thermal power ~5%

3 Hot Standby Keff < 0.99 and average coolant temperature ~350°F 4 Hot Shutdown Keff < 0.99 and average coolant temperature 350°F > Tavg > 200 °F 5 Cold Shutdown Keff < 0.99 and average coolant temperature ~200°F 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 18-1 January, 2018

3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Coordinator/EDF 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/EDF 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 18-1 January, 2018

3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 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/EOF 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/EOF 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 18-1 January, 2018

3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at

, two different units, a Site Area Emergency should be declared.

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

If two Alert EALs are met, whether at one unit or at two 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/EOF 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 ECL 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 18-1 January, 2018

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 ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response - In instances 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 exan,ple:

An ATWS 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.

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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/EDF Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulfing 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 NRC 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 NRC is discussed in NUREG-1022 (ref. 4.1.3).

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4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.

4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50. 72 and 50. 73 4.1.4 10 § CFR 50. 72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50. 73 License Event Report System 4.1.6 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 4.1.1 O 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 fy1atrix 4.2.3 MNS EAL Matrix D-14 Revision 18-1 January, 2018

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 (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE)

Alert

Site Area Emergency (SAE)

General Emergency (GE)

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EPA PAGs Environment 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 inspection to determine if the attributes of an explosion are present.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

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

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

Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

General Emergency Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile 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.

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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).

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)

Personnehaccess 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.

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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 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.

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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.

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5.2 Abbreviations/Acronyms

~F ....................................................................................................... Degrees Fahrenheit 0

- - - - * " . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Degrees 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 ...............................................................................................................Direct Current EAL ............................................................................................. Emergency Action Level EC ............................................................................................... Emergency Coordinator ECCS ............................................................................ Emergency Core Cooling System ECL. ................................................................................. Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ... '. ........................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency 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 Kett ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER ................................................................................................ Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident D-20 Revision 18-1 January, 2018

LWR................................................................................................... Light Water Reactor MPC....... ... ... ..... ........... ...... Maximum Permissible Concentration/Multi-Purpose Canister MSIV.......................................................................................Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line mR, mRem, mrem, mREM .............................................. mi Iii-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 QBE ....................................................................... :.............. 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/ 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 18-1 January, 2018

SI .............................................................................................................. Safety Injection SLC ................................................................................. Selected Licensee Commitment SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ SeniorReactorOperator SSF .......................................................................................... Standby Shutdown Facility TEDE ............................................................................... Total Effective Dose Equivalent TOAF .................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center WOG .................................................................................. Westinghouse Owners Group D-22 Revision 18-1 January, 2018

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 NEI 99-01 Rev. 6 Example EAL IC EAL RU1.1 ~U1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 D-23 Revision 18-1 January, 2018

MNS NEI 99-01 Rev. 6 Example EAL IC EAL CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1, 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 3 CG1.1 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1 HU1.2 HU1 2 HU1.3 HU1 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 D-24 Revision 18-1 January, 2018

MNS NEI 99-01 Rev. 6 Example EAL IC EAL HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1 HA1.2 HA1 2 HA5.1 HA5 1 HA6.1 HA6 1 HA1.1

HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 D-25 Revision 18-1 January, 2018

MNS NEI 99-01 Rev. 6 Example EAL IC EAL SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SGB 1 EU1.1 E-HU1 1 D-26 Revision 18-1 January, 2018

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 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category R-Abnormal Rad Release/ Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

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

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ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the 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: 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.

Note 3: 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.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE U)

Unit Vent Noble Gas Low 1(2)EMF36L -- --- 4.85E+6 cpm 3.10E+3 cpm

J 0

31 Unit Vent Noble Gas High I'll 1(2)EMF36H 2.61E+4 cpm 2.61E+3 cpm 2.70E+2 cpm ---

C)

Liquid Waste Effluent Line High EMF49H --- -- -- 2.15E+2 cpm "Cl

3 CVUCDT High C"

1(2)EMF44H -- - - 4.29E+2 cpm

J Mode Applicability:

All Definition(s):

None Basis:

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Liquid Releases 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 low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

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

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

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

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

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.

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ATTACHMENT 1 EAL Bases MNS Basis Reference(s}:

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the 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 for ~ 60 min. (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):

None Basis:

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

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

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone.

The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

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

D-32 Revision 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 1O 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: 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.

Note 3: 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.

Note 4 The pre-calculated effluent monitor values presented in EALs RA1 .1, R.S1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point. Monitor GE SAE Alert UE en Unit Vent Noble Gas Low 1(2)EMF36L --- -- 4.85E+6 cpm 3.10E+3 cpm

I 0

GI en Unit Vent Noble Gas High ca C) 1(2)EMF36H 2.61E+4cpm 2.61E+3 cpm 2.70E+2 cpm --

Liquid Waste Effluent Line High EMF49H -- - - 2.15E+2 cpm "ti

s CVUCDT High 1(2)EMF44H --- --- --- 4.29E+2 cpm tr

i Mode Applicability:

All Definition(s):

None D-34 Revision 18-1 January, 2018'

ATTACHMENT 1 EAL Bases Basis:

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 AA 1 D-35 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.2 Alert 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, 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 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 18-1 January, 2018

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

2. NEI 99-01 AA 1 D-37 Revision 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (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 retease 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 18-1 January, 2018

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 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 1O mrem TEDE or 50 mrem thyroid COE EAL:

RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for 2:: 60 min.

Analyses of field survey samples indicate thyroid COE > 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:

AD-EP-MNS-0203 MNS Site Specific Field Monitoring and AD-EP-ALL-0203 Field Monitoring During Declared Emergency provide 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases alone. The inclusion of both piant 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.

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

MNS Basis Reference(s):

1. AD-EP-MNS-0203 MNS Site Specific Field Monitoring

2. AD-EP-ALL-0203 Field Monitoring During Declared Emergency.

3. NEI gg;.Q1 AA1 D-41 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE 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: 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.

Note 3: ILthe 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.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE U)

Unit Vent Noble Gas Low 1(2)EMF36L -- -- 4.85E+6 cpm 3.10E+3 cpm

r 0

GI U)

('Cl C)

Unit Vent Noble Gas High 1(2)EMF36H 2.61E+4 cpm 2.61E+3 cpm 2.70E+2 cpm ---

Liquid Waste Effluent Line High EMF49H --- --- - 2.15E+2 cpm "Cl

3 CVUCDT High C"

1(2)EMF44H --- --- --- 4.29E+2 cpm

i Mode Applicability:

All Definition(s):

None D-42 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Basis:

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 classifY,ing-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 EPAPAG 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid 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 greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone.

The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RG1.

MNS Basis Reference(s):

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

2. NEI 99-01 AS1 D-45 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid 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 for~ 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/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:

AD-EP-MNS-0203 MNS Site Specific Field Monitoring ahd AD-EP-ALL-0203 Field Monitoring During Declared Emergency provide 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases 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.

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

MNS Basis Reference(s):

1. AD-EP-MNS-0203 MNS Site Specific Field Monitoring Information

2. AD-EP-ALL-0203 Field Monitoring During Declared Emergency.

3. NEI 99-01 AS1 D-47 Revision 18-1 January,2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid 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: 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.

Note 3: lfthe 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.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE C/J Unit Vent Noble Gas Low 1(2)EMF36L -- - 4.85E+6 cpm 3.10E+3 cpm

s 0

5!CII C)

Unit Vent Noble Gas High 1(2)EMF36H 2.61E+4 cpm 2.61E+3 cpm

2.70E+2 cpm ---

Liquid Waste Effluent Line High EMF49H -- --- --- 2.15E+2 cpm "Cl

3 CVUCDTHigh C'

1(2)EMF44H -- --- --- 4.29E+2 cpm

J Mode Applicability:

All Definition(s):

None D-48 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Basis:

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid 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 RA 1.1, 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.

a Classification based on effluent monitor readings assumes that 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

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

2. NEI 99-01 AG1 D-51 Revision 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 -- Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid 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/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:

AD-EP-MNS-0203 MNS Site Specific Field Monitoring and AD-EP-ALL-0203 Field Monitoring During Declared Emergency provide 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 18-1 January, 2018 L _ _ ___ _

ATTACHMENT 1 EAL Bases 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.

MNS Basis Reference(s):

1. AD-EP-MNS-0203 MNS Site Specific Field Monitoring

2. AD-EP-ALL-0203 Field Monitoring During Declared Emergency.

3. NEI 99-01 AG1 D-53 Revision 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.1 Alert 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 18-1 January, 2018

. __ J

ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND A Trip 2 radiation alarm on any of the following radiation monitor indications:

1EMF17 (2EMF4) Spent Fuel Building Refueling Bridge

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

1EMF42 (2EMF42) Fuel Building Ventilation

1EMF39 (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 cir 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Escalation of the emergency would be based on either Recognition Category R or C ICs.

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/0/8/1003/008 Determination of Radiation Monitor Setpoints (EMFs)

3. NEI 99-01 AA2 D-59 Revision 18-1 January,, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Spent fuel pool level s -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 1O 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 addresse,s 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 ICs.

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 18-1 January, 2018

ATIACHMENT 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 i

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 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Spent fuel pool level~ -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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to> -25 ft. (746 ft. ele.) (KFP5350 or NVPG6530) for ~ 60 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:

All Definition(s):

None Basis:

Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of ]dentifying 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.) (7 45 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.1 Alert Dose 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 18-1 January, 2018

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 ICs.

MNS Basis Reference(s):

1. UFSAR Table 12-11 Area Radiation Monitors

2. NEI 99-01 AA3 D-67 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Conditi(?n: 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 P/C, RHole, near 1Nl-185, ABPC thru CAD Door, FF59

~uxiliary 716' 4 Outside CAD 212 800 (1 EMXA) 820 {2EMXA) 3,4

~uxiliary 750' 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 (1 ETB) 716 (2ETB) 3,4 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).

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 18-1 January, 2018

ATTACHMENT1 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 ifextraordinary 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 ICs.

D-69 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases

2. NEI 99-01 AA3 D-70 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown / 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: .L 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.

Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases

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

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 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 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):

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 18-1 January, 2018

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 9R-01 CU1 D-74 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: 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 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 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 18-1 January, 2018

ATIACHMENT 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/5500/1 O 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory EAL:

CA1.1 Alert 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 CA 1

D-77 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory EAL:

CA1.2 Alert NCS water level cannot be monitored for~ 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

'limithas been exceeded, or will likely be exceeded.

fable C-6 Sumps/Tanks

NCDT

PRT

CFAEsump

ND/NSsump

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 18-1

January, 2018

ATIACHMENT 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/4150/001 D Identifying NC System Leakage

3. NEI 99-01 CA1 D-79

. Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 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 Sumps/Tanks

NCDT

PRT

CFAEsump

ND/NSsump

RHT

WOT

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.

D-80 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases 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 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 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 18-1 January, 2018

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)/Af5500/10 NC System Leakage Within the Capacity of Both NV Pumps

2. PT/1 (2)/Af4150/001 D Identifying NC System Leakage 3.

NEI 99-01 CS1 Category: C - Cold Shutdown / Refueling System. Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1 .*1.i 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 1EMF16 (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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Table C-6 Sumps/Tanks

NCDT

PRT

CFAEsump

ND/NSsump

RHT

WOT

WEFT

SRST fable C-1 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 C are 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 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 D-83 Revision 18-1 January, 2018

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 melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If 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 re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to.support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment D-84 Revision 18-1 January, 2018

ATTACHMENT 1 l=AL 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 levehdropping 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 NU MARC 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 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 for =!! 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit. has been exceeded, or will likely be exceeded.

fable C-2 AC Power Sources

>ffsite:

ATC (Train A)

SATA (Train A)

ATD (Train 8)

SATB (Train 8)

>nsite:

DIG 1(2) A (Train A)

DIG 1(2) 8 (Train 8)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, NM - No Mode D-86 Revision 18-1 January, 2018

ATIACHMENT 1 EAL Bases Definition(s):

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

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

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 6.9KV/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) 8) 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 18-1 '

January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer EAL:

CA2.1 Alert 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 18-1 January, 2018

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.:O.O EP/1 (2)/A/5000/ECA-O.O Loss of All AC Power

4. NEI 99-01 CA2 D-90 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: UNPLANNED increase in NCS temperature EAL:

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 18-1 January, 2018

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 CA 1 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3- NCS Temperature Initiating Condition: UNPLANNED increase in NCS temperature EAL:

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 18-1 January, 2018

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):

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:*

CA3.1 Alert 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: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

Note 9: 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 Containment Closure NCS Status Heat-up Duration Status Intact (but not reduced N/A 60 min.*

inventory)

Not intact established 20 min.*

OR not established 0 min.

At reduced inventory

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 C are met.

D-95 Revision 18-1 January, 2018

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 (1 NCLP5122 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 heat-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 18-1 January, 2018

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. 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event

< 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 1E 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 1E 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 18-1 January, 2018

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 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 CA 1 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/A/5500/15 Loss of Vital or Aux Control Power

3. MNS UFSAR Section 8.0 Electrical Power

4. NEI 99-01 CU4 D-99 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CUS.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 Onsite ORO NRC Public Address X Internal Telephones X Onsite Radios X

DEMNET X Offsite Radio System X Commercial Telephones X X NRC Emergency Telephone System (ETS) X Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, NM - No Mode Definition(s):

None Basis:

Onsite/offsite communications include one or more of the systems listed in Table C-4 (ref. 1).

D-100 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases 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. .,

NRC Emergency Telephone System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC 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 NRC 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 18-1 January, 2018

ATIACHMENT 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 NRC.

This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to notify all 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-102 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 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 Table 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 Mode Applicability:

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 18-1 January, 2018

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 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

VISIBLE DAMAGE - Damage to a 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 18-1 January, 2018

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 18-1 January, 2018

ATIACHMENT 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.

4.Fire 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 18-1 January, 2018

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/EDF Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Coordinator/EDF Director judgment.

D-107 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by 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 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 ~ostile 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 18-1 January, 2018

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 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-109 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: _ 1 - Security Initiating Condition: Confirm,ed SECURITY CONDITION or threat EAL:

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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases NRC. 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 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-113 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by 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 s~curity-related event (ref. 2, 3, 4).

D-114 Revision 18-1 January, 2018

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/Af5500/47 Security Events

3. AP/O/Af5500/48 Extensive Damage Mitigation

4. NEI 99-01 HA1 D-115 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H- Hazards Subcategory: 1 - Security .

Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREAor airborne attack threat within 30 minutes EAL:

HA1.2 Alert 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 eq111ipment, 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 18-1 January, 2018 I

J

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 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated incaccordance 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Hostile Action within the Protected Area EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by 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 eqllipment, 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 18-1 January, 2018

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'\,vould 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Basis:

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 threatfocation. 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than QBE levels EAL:

HU2.1 Unusual Event Seismic event> OBE 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 (QBE) 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 QBE 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.gov/eqcenter/

D-122 Revision 18-1 January, 2018

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 1MC9. 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 safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk- downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

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

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

All Definition(s):

PROTECTED AREA - 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 ICs in Recognition Categories R, F, S or C.

MNS Basis Reference(s):

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

2. NEI 99-01 HU3 D-124 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode Mode Applicability:

All Definition(s):

I 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 18-1 January, 2018

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 ICs 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

Mode Applicability:

All Definition(s):

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

PROTECTED AREA - 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 ICs in Recognition Categories R, F, Sor C.

MNS Basis Reference(s):

1. NEI 99-01 HU3 D-127 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

All Definition(s):

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 ICs in Recognition Categories R, F, Sor C.

MNS Basis Reference(s):

1. NEI 99-01 HU3 D-128 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

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

Receipt of multiple (more than 1) fire 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/EOF 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 alarm D-129 Revision 18-1 January, 2018

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 subsequentverification 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire 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 t: The Emergency Coordinator/EOF 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 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 18-1 January, 2018

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/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.

a This EAL addresses receipt of 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 sing'le.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 18-1 January, 2018

ATIACHMENT 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the 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 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.

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 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 Unit 2 Room/Area Modes P/C, RHole, near 1Nl-185, ABPC thru CAD Door, FF59 Auxiliary 716' 4 Outside CAD 212 800 (1EMXA) 820 (2EMXA) 3,4 Auxiliary 750' 803 (1 ETA) 805 (2ETA) 3,4 702 (Elec. Pene.) 713 (Elec. Pene.) 3 Auxiliary 733' 722 (1 EMXB-1) 724 (2EMXB-1) 3,4 705 (1 ETB) 716 (2ETB) 3,4 Mode Applicability:

3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

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 18-1 January, 2018

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 18-1 January, 2018

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 ICs.

NOTE: JC HA5 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 JC HA5 mode applicability is required.

MNS Basis Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases

2. NEL99-01 HA5 D-138 Revision 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in 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 adjacentto 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases 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 HA6 D-140 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in 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 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).

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Coordinator/EOF 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 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 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/EOF 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/EOF 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 18-1 January, 2018

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/EOF 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 8.2 Emergency Coordinator

2. NEI 99-01 HU?

D-144 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator/EOF Director warrant declaration of an Alert EAL:

HA7.1 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):

HOST/LE 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 wide-ranging responsibilities associated with managing a major emergency (ref.1 ).

D-145 Revision 18-1 January, 2018

ATIACHMENT 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 B.2 Emergency Coordinator

2. NEI 99-01 HA7 D-146 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator/EOF Director warrant declaration of a Site Area Emergency EAL:

HS7.1 Site Area 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 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/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 D-147 Revision 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator/EOF Director warrant declaration of a General Emergency EAL:

HG7.1 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/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 wide-ranging responsibilities associated with managing a major emergency (ref. 1).

D-149 Revision 18-1 January, 2018 L

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 believed 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 8 On-Site Emergency Organization Section 8.2 Emergency Coordinator

2. NEI 99-01 HG7 D-150 Revision 18-1 January, 2018

ATIACHMENT 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all offsite AC power capability to essential buses for 15 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.

Table S-1 AC Power Sources Dffsite:

ATC (Train A)

SATA (Train A)

ATD (Train 8)

SATB (Train 8)

Dnsite:

DIG 1(2) A (Train A)

DIG 1(2) 8 (Train 8)

Mode Applicability:

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 18-1 January, 2018

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 6.9KV/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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 offsite power.

Escalation of the emergency classification level would be via IC SA1.

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 SU1 D-154 Revision 18-1

January, 2018

ATTACHMENT 1 EAL Bases Category: S- System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:

SA1.1 Alert AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and 1(2)ETB reduced to a single power source for ~ 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table 5-1 AC Power Sources Dffsite:

ATC (Train A)

SATA (Train A)

ATD (Train 8)

SATB (Train B)

>nsite:

DIG 1(2) A (Train A)

DIG 1(2) B (Train B)

Mode Applicability:

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 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

D-155 Revision 18-1 January, 2018

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.

Basis:

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 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 18-1 January, 2018

ATIACHMENT 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-O.O Loss of All AC Power

4. NEI 99-01 SA 1 D-157 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to 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/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:

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 ET8 (Train 8) (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 6.9KV/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 & SAT8). 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) 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 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months S80 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely (Note 1)

Core Cooling RED PATH conditions met 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:

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 6.9KV/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).

D-160 Revision 18-1 January, 2018

- - - - - - --------i 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 18-1 January, 2018 l

ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

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 Loss of Electrical Power *

5. ECA-0.0 EP/1 (2)/A/5000/ECA-0.0 Loss of All AC Power

6. NEI 99-01 SG1 D-162 Revision 18-1

. January, 2018

ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all essential AC and vital DC power sources for 15 minutes or longer 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Definition(s):

None Basis:

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 6.9KV/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 (DIG 1(2) A & D/G 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 autpmatically 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SBO event (ref. 1).

The 125 VDC electrical power system consists of two independent and redundant safety related Class 1E 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 1E 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 18-1 January, 2018

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 SG8 D-165 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:

SS2.1 Site Area Emergency 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 1E 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 1E 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 ICs RG1, FG1 or SG1.

D-166 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

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 SS8 D-167 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for ~ 15 min. (Note 1)

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

fable S-2 Safety System Parameters

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 18-1 January, 2018

ATIACHMENT 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 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 leve*1 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)/A/6100/SD-2 Cooldown to 400 Degrees F

3. NEI 99-01 SU2 D-169 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for ~ 15 min. (Note 1)

AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Safety System Parameters

Reactor power

NCS level

NCS pressure

Core exit T/C temperature

Level in at least one S/G

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 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown D-170 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases 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).

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 18-1 January, 2018

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 ICs FS1 or IC RS1 MNS Basis Reference(s):

1. UFSAR Section 7.5 Safety-Related Display Instrumentation

2. OP/1(2)/A/6100/SD-2 Cooldown to 400 Degrees F

3. NEI 99-01 SA2 D-172 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 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 - Powe'r Operations, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The specific iodine activity is limited to s 1.0 µCi/gm Dose Equivalent 1-131 for> 48 hrs. or s 60 µCi/gm Dose Equivalent 1-131 instantaneous.* The specific Xe-133 activity is limited to s 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 5 - NCS Leakage Initiating Condition: NCS leakage for 15 minutes or longer EAL:

SUS.1 Unusual Event NCS unidentified or pressure boundary leakage > 10 gpm for ~ 15 min.

OR NCS identified leakage > 25 gpm for~ 15 min.

OR Leakage from the NCS to a location outside containment > 25 gpm 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:

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 18-1 January, 2018

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.

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 ICs 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by 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 Definition(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 18-1 January,2018

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 (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) 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 (2)/A/5000/FR-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 18-1 January, 2018

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 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 cJown 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 SU5 D-178 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.2 Unusual Event A manual trip 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 18-1 January, 2018

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, arid 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 shpuld be applied.

D-180 Revision 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:

SA6.1 Alert An automatic or manual trip 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 (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) 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core coofing or NCS heat removal EAL:

556.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power

~5%

AND All actions to shut down the reactor are not successful as indicated by reactor power

~5%

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)/N5000/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 18-1 January, 2018

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 SS5 D-185 Revision 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 ORO communication methods OR Loss of all Table S-4 NRC communication methods Table 5-4 Communication Methods System Onsite ORO NRC Public Address X Internal Telephones X

Onsite Radios X

DEMNET X Offsite Radio System X Commercial Telephones X X NRC Emergency Telephone System (ETS) X Mode Applicability:

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

None D-186 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Basis:

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 cap~bility 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.

NRC Emergency Telephone System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC 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 NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.

D-187 Revision 18-1 January, 2018

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 on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locatic;ms, 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control.

EAL:

SUS.1 Unusual Event EITHER:

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 2:: 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 1O 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 18-1 January, 2018 J

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 CARF 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 challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

The second condition addresses a condition where containment pressure is greater than the setpoint atwhich 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 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 degr~ded 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 fable S-5 Hazardous Events

Seismic event (earthquake)

Internal or external FLOODING event

High winds or tornado strike

FIRE

EXPLOSION

Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability:

1 - Power 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 18-1 January, 2018

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 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 ir, 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a 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, HA 1.1 and HS1 .1 Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.

D-194 Revision 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: E- ISFSI Sub-category: 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 (excludes air side of the cask

340 mrem/hr (gamma) or 40 mrem/hr (neutron) on the sides of the radial neutron inleUoutlet ports)

100 mrem/hr shield

10 mrem/hr (neutron +

560 mrem/hr (gamma) or 280 mrem/hr (neutron) on the side gamma) on the top (neutron) on the side surfaces above the of the cask of the cask (excludes air radial neutron shield region inleUoutlet ports)

200 mrem/hr

220 mrem/hr (gamma) or 400 mrem/hr (neutron+

(neutron) on the side surfaces below the

900 mrem/hr gamma) at air radial neutron shield region (neutron+ gamma) inlets and outlets on the top of the cask (excludes air inleUoutlet 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Basis:

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, off-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 18-1 January, 2018

ATTACHMENT 1 EAL Bases MNS Basis Reference(s):

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 18-1 January, 2018

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 <FC): 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

~Isa 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases

Unusual 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 as-designed/expected operation of a relief valve is not considered to be NCS leakage.

Atthe 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: NIA Initiating Condition: Any loss or any potential loss of either Fuel Clad or NCS EAL:

FA1.1 Alert 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 18-1 January, 2018

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1.1 Site Area Emergency Loss OR potential loss of any two barriers (Table F-1)

Mode Applicability:

1 - Power 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 (i.e., loss - loss)

One barrier loss and a second barrier potential loss (i.e., loss - potential loss)

One barrier potential loss and a second barrier potential loss (i.e., potential loss -

potential loss)

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and 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 18-1 January, 2018

ATIACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A lnitjating 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 18-1 January, 2018

ATIACHMENT2 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. GMT Radiation/ NCS Activity D. GMT 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 "GMT 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 18-1 January, 2018

ATIACHMENT2 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 8, ... , E.

D-204.

Revision 18-1 January, 2018

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 Containment (CMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. An automatic or manual ECCS 1. Operation of a standby charging pump is required by EITHER:

A (SI) actuation required by NCS or SG Tube None None EITHER:

UNISOLABLE NCS .

UNISOLABLE NCS leakage SG tube leakage

1. A leaking or RUPTUREO SG is FAULTED outside of containment None Leakage

. leakage SG tube RUPTURE

2. Integrity-RED PATH conditions met

1. Core Cooling-ORANGE PATH conditions met 1. Core Cooling-RED PATH 8 1. Heat Sink-RED PATH conditions met conditions met

1. Core Cooling-RED PATH 2. Heat Sink-RED PATH conditions Inadequate met None None IAND conditions met AND Heat AND Restoration procedures not Heat sink is required Removal effective within 15 min. (Note 1)

Heat sink is required C 1. EMF51NB > Table F-2 column

1. EMF51NB > Table F-2 column CMT "FC Loss" 1. EMF51NB > Table F-2 column None "NCS Loss" None None Radiation 2. Dose equivalent 1-131 coolant "GMT Potential Loss"

/NCS activity> 300 µCi/gm Activity

1. Containment isolation is required AND EITHER:

. Containment integrity has

1. Containment-RED Path conditions met D been lost based on Emergency

2. Containment hydrogen concentration

>6%

None None None None Coordinator/EOF Director CMT 3. Containment pressure > 3 psig judgment Integrity or Bypass . UNISOLABLE pathway from Containment to the environment with EITHER a failure of both trains of NS OR failure of both trains of VX-CARF for~ 15 min.

exists (Notes 1, 1O)

2. Indications of NCS leakage outside of containment

1. Any condition in the opinion of E the Emergency Coordinalor/EOF

1. Any condition in the opinion of the Emergency Coordinator/EOF Director

1. Any condition in the opinion of lhe Emergency Coordinator/EOF Director Director that indicates loss of the EC ~hat indicates potential loss of the fuel hat indicates loss of the NCS barrier that indicates potential loss of the NCS that indicates loss of the containment that indicates potential loss of the uel clad barrier clad barrier barrier barrier containment barrier Judgment D-205 Revision 18-1 January, 2018

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:

None D-206 Revision 18-1 January, 2018

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 18-1 January, 2018

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 18-1 January, 2018

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:

Critical 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 sufficient 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 Revision 18-1 January, 2018

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 AND 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 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 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 (34 7°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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Loss Threshold:

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. EMFf?1 & 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 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 18-1 January, 2018

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 condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for 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 / RCS Activity Fuel Clad Loss 3.A D-213 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / 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/ Containment Radiation.

MNS Basis Reference(s):

1. RP/O/A/5700/019 Core Damage Assessment

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.8 D-214 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. GMT Radiation / NCS Activity Degradation Threat: Potential Loss Threshold:

D-215 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

D-216 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

D-217 Revision 18-1 January, 2018

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/EDF 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 EDPs. The Emergency Coordinator/EDF Director 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 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 18-1 January, 2018

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 Coordihator/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 ATWS 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 18-1 January, 2018

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):

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

RUPTURE,::. The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Basis:

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 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. 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 18-1 January, 2018

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:

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 18-1 January, 2018

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 18-1 January, 2018

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 1.8 D-223 Revision 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: 8. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. Heat Sink-RED path conditions met AND Heat sink is required Definition(s):

None Basis:

In combination with FC Potential Loss 8.2, 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 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 T hot 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 18-1 January, 2018

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 8.2; 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 2.8 D-226 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT 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 S/D (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 1E8 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 18-1 January, 2018

ATIACHMENT2 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/ Containment Radiation.

MNS Basis Reference(s):

1. RP/O/A/5700/019 Core Damage Assessment

2. NEI 99-01 CMT Radiation / RCS Activity NCS Loss 3.A D-228 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CMT Radiation/ NCS Activity Degradation Threat: Potential Loss Threshold:

D-229 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

D-230 Revision 18-1 January, 2018

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 18-1 January, 2018

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 18-1 January,2018

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 Eme~gency 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 "immine*nt" 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 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 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 18-1 January, 2018

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):

FAULTED - The term applied to a steam generator that has a steam-leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficientmagnitude 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.

FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e.,

NCS activity values) and IC SUS 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 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.

Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (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-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires operation of a standby charging .

(makeup) pump (NCS Barrier Potential Loss) ~~~ -~rea Emergency per Alert per FA1 .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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

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 subsequent challenge of the Containment Barrier.

The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency 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 thatfunction 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/NCS Activity Degradation Threat: Loss Threshold:

D-239 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/NCS Activity Degradation Threat: Potential Loss Threshold:

1. EMF51A/B > Table F-2 column "CMT Potential Loss" Table F-2 Containment Radiation - R/hr (EMF51A & B)

Time After S/D (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 CMT 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 18-1 January, 2018

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 / RCS Activity Containment Potential Loss 3.A D-241 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND 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 accident conditions, 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 FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

D-242 Revision 18-1 January, 2018

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 GMT Integrity or Bypass Containment Loss 4.A D-243 Revision 18-1 January, 2018

ATIACHMENT2 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 18-1 January, 2018

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 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Inside Reactor Building Damper 1------------1 I I

Area

Monitor ,

1_ - - - - - - - - - - _I I

Process

Monitor I I

- - - - - - - - - - ,I RCP D-246 Revision 18-1 January, 2018

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 18-1 January, 2018

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 4.8 D-248 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT 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 for~ 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 1O: 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 contain'ment 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 18-1 January, 2018

ATIACHMENT2 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 CMT Integrity or Bypass Containment Potential Loss 4.C D-250 Revision 18-1 January, 2018

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

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 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 Loss 6.A D-251 Revision 18-1 January, 2018

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 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 Potential Loss 6.A D-252 Revision 18-1 January, 2018

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases

Background

NEI 99-01 Revision 6 ICs 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 18-1 January, 2018

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:

OP/1&2/A/6100/003, Perform OP/1/A/6100/SD 1 N/A N/A No Enclosure 4.2, (Prepare For Cooldown).

Step 3.8.4.1 OP/1&2/A/6100/003 Perform NC System degas N/A N/A No

, 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 Transfer Yard 1 No Enclosure 4.2, Transformer Cooling Steps 3.8.8.1 & Groups.

3.8.9.1 OP/1&2/A/6100/003, Perform Main Steam Main Steam Doghouses No Enclosure 4.2, Safety Valve testing.

Step 3.8.14 OP/1&2/A/6100/003, Check transfer of Aux Turbine Bldg. Basement 1 No Enclosure 4.2, Steam from C htr Bleed to (739') North Wall Step 3.8.17.1 Main Steam (Close 1SP-1 (Main Steam to 1A CF Pump Turb lsol) and 1SP-2 (Main Steam to 1B CF Pump Turb lsol).1AS-11 ).

OP/1&2/A/6100/003 Stop G HOT Pumps per Turbine Bldg. Basement No

, Enclosure 4.2, OP/1 /B/6250/004 (739') West Wall Step 3.8.21 (Feedwater Heater Vents, Drains, and Bleed System).

OP/1&2/A/6100/003, Stop C HDT Pumps per Turbine Bldg. Basement 1 No Enclosure 4.2, OP/1 /B/6250/004 (739') HP Heater Panel Step 3.8.23 (Feedwater Heater Vents, Drains, and Bleed System).

OP/1&2/A/6100/003, Transfer of Aux Steam to Service Bldg. (739') or No Enclosure 4.2, Unit 2 or Aux Electric Auxiliary Boiler Room 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 Turbine Bldg. Mezz (760') No Enclosure 4.2, to 1A CF Pump Turb lsol) at CF Pumps Step 3.12.7 and 1SP-2 (Main Steam to 1B CF Pump Turb lsol).

D-254 Revision 18-1 January,2018

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases OP/1&2/A/6100/003, Shutdown MG Sets per MG Set Room (767') 3 No .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 3 No .2, Layup Chemical addition and TB Basement (739'}

Step 3.14.4 {Chemistry).

OP/1&2/A/6100/003, Begin performance of Inside Containment 3 No .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 3 No 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/S0- After required amount of Aux. Bldg. (NM Lab 716') 3 No 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/S0- After required amount of Aux. Bldg. (NM Lab 716') 3 No 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/S0- After required amount of Aux. Bldg. (NM Lab 716') 3 No 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/S0- After required amount of Aux. Bldg. (NM Lab 716') 3 No 1, Step 3.6. 7 boron is added for Counting Room (767')

Refueling Boron Concentration, Primary Chemistry samples NC S stem.

OP/1&2/A/6100/SO- Have Radwaste align Aux. Bldg. (716') 1-3 No D-255 Revision 18-1 January, 2018

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases V v,* *'_,,, < '/' ,,._

'- .'. :.. *.~;r1.:lfJk -~-;1~::J~J):1: :11;: ~~t~!?;\ ,

M~~i~f

~nd,,~W,) ~~ilding/E:f~y.at.io~~o~pj;. *.

, *.,,- ** >. ., r,,~ '<"<</::.: , -.._;:p'

>;.'.}'

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') 1-3 No 10, Step 3.6.2 PRT Degas per Radwaste Area OP/O/A/6200/518 (Waste Gas Operation).

OP/1 &2/A/61 OO/SD- Radwaste performs NC Aux. Bldg. (716') 1-3 No 10, Step 3.7.1 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 Aux. Bldg. (716') 1-3 No 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 Aux. Bldg. (716') 1-3 No 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. Aux. Bldg .. (733') BAT 3 No 2, Step 3.2.3 Area OP/1&2/A/6100/SD- When less than 1000 psig, Aux. Bldg. (733') Electrical 3. Yes 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 3 No 2, Enclosure 4.2, adjust NC Pump Seal Water Outside VCT Room Step 3.2.3.2 Injection Throttles to 8-10 gpm.

OP/1&2/A/6100/SD- Radwaste crossties BATs. Aux. Bldg. (733') BAT 3 No 4, Step 3.2.3 Area OP/1&2/A/6100/SD- Perform plant shutdown ETA, ETB, Aux. Bldg. 3 Yes 4, Steps 3.11.1, tagging. (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 3 No 4, Enclosure 4.2, adjust NC Pump Seal Water Outside VCT Room Step 3.10.2 Injection Throttles to 8-10 gpm.

D-256 Revision 18-1 January, 2018

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases OP/1&2/N6100/SD- Secondary Chemistry to Chemistry Lab and 4 No 4, Step 3.15 check S/G sulfate meets Turbine Bldg. (739') North chemistry criteria for Wall continued cooldown.

OP/1&2/N6100/SO- Stroke time testing of Inside Containment 4 No 10, Step 3.8 PORVs per PT/1/N4151/005 (NC Valve Stroke Timing Test Using Air). Enclosures 13.4, 13.5, 13.6 OP/1&2/N6100/SO- Rack out and tag one NV ETA (750'), ETB (733') 4 Yes 10, Step 3.11.1, and both NI Pumps per 3.11 .2, 3.11 .3 OP/O/N6350/008 (Operation of Station Breakers).

OP/1&2/N6100/SO- Tag out PD Pump at 1MXK- ux. Bldg. (750') North 4 Yes 10, Step 3.11 .6 F2C (Reciprocating Charging End Pump No 1).

OP/1&2/N6100/SO- If LTOP vent requirements are Inside Containment 4 No 10, Step 3.12 to be satisfied by securing 1NC-36B (Pzr PORV) open, Maintenance gags 1NC-36B (Pzr PORV).

Unlock and close 1/2ND- P/C, RHole, near 1Nl-185, 119 (1/2 A ND ECCS Outside CAD 212 (716')

Sump Suction Relief Inlet BPC thru CAD Door, lsol #2).

FF59 (716')

OP/1 &2/N61 OO/SD- Perform PT/1/N4206/030 ux Building Pipechase 4 No 6A(B), Enclosure (Draining ECCS Sump Piping (716')

4.1, Step 3.6 Drain Reservoir Train A),

Enclosure 13.1 (Draining ECCS Sump Piping Drain Reservoir rain A in Modes 1 4) and PT/1/N4206/031 (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/N6100/SD- Monitor and shift NC System ux Building (716'/733') at 4 No 6A(B), Encl. 4 .. 1, Filters on high DP. NC Filters Room Step 3.8 D-257 Revision 18-1 January, 2018

ATIACHMENT3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases OP/1 &2/A/61 OO/SD- De-energize 1/2ND-68A (A ETA Aux. Bldg. (750') 4 Yes 6A(B), Encl. 4.2 ND Pump & A Hx Miniflow)

Step 3.7 in the open position at 1/2EMXA-F12B (1/2ND-68A).

OP/1 &2/A/61 OO/SD- Adjust flow through Cation Aux Building (750') over 4 No 6A{B), Step 3.31.1 Bed Demineralizer per Demin Pits OP/1/A/6200/001 D

{Chemical and Volume Control System Dem ineralizers ).

OP/1&2/A/6100/SO- De-energize 1/2ND-67B (B ETB Aux. Bldg. (733') 4 Yes 6, Step 3.4.3 ND Pump & B Hx Miniflow) in the open position at 1/2EMXB1-2C (1/2ND-67B).

Table R-2 & H-2 Results Table R-2/H-2 Safe Operation & Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area Unit 2 Room/Area Modes P/C,RHole, near 1Nl-185, ABPC thru CAD Door, FF59 4 Auxiliary 716' Outside CAD 212 800 (1EMXA) 820 (2EMXA) 3,4 Auxiliary 750' 803 (1 ETA) 805 {2ETA) 3,4 702 {Elec. Pene.) 713 {Elec. Pene.) 3 Auxiliary 733' 722 (1 EMXB-1) 724 (2EMXB-1) 3,4 705 (1 ETB) 716 {2ETB) 3,4 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 D-258 Revision 18-1 January, 2018

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases D-259 Revision 18-1 January, 2018

G. Public Education and Information Information will be made available to the public on a yearly basis concerning notification of a nuclear plant emergency and the response that will be required from the public sector.

G.1/G.2 Public Education and Information Program On an annual basis, the licensee will update and distribute to residents within the plume exposure pathway emergency-planning zone, emergency-planning information concerning McGuire Nuclear Station. It will provide educational information on radiation, emergency planning contacts, protective actions, primary emergency alert system radio stations, evacuation routes, pick-up points for school children, reception centers, and information for residents with special needs. Local telephone numbers to call with questions will also be listed.

Public information for the transient population includes lake-access signs and emergency planning information. Transient locations will be identified by the Emergency Planning Manager/designee, the site Nuclear Communications staff, and state and county emergency management officials. These locations may include but are not limited to motels, hotels, marinas, and lake access (signs).

The list of transient locations will be reviewed quarterly and updated as needed. Locations will be contacted periodically to ensure adequate copies of materials are available.

G.3.a Communications - Location and Contacts During a drill or emergency, public information at McGuire Nuclear Station will be coordinated and disseminated through the on-site media center located on Hagers Ferry Rd., Huntersville, N.C. or the Joint Information Center (JIC) located in the Energy Center at 526 South Church Street, Charlotte, N. C.

During initial stages of an emergency situation, response to media questions relative to plant status will be provided at the on-site media center. The Charlotte media center, also located in the Energy Center, will be activated as needed. The news release will indicate the location of the primary media center. A company spokesperson will be designated as the primary contact for the news media.

If the Emergency Operations Facility (EOF) is not activated, the normal Duke Energy news release process is followed. If the EOF is activated, then the Joint Information Center (JIC) implements I

procedures for gathering and disseminating information..

G.3.b Communications - Media Center In a nuclear plant emergency, the licensee relies on the news media to provide prompt, accurate information to local residents and the public. To provide ready access to current information on plant status, a media center is promptly established. An on-site media center will provide space for a limited number of media. A larger media center, located in the Energy Center at 526 South Church Street, Charlotte, NC (near the EOF) can be activated as needed to support additional media.

G.4.a Company Spokesperson A company spokesperson will provide plant status and company information during scheduled news conferences and media briefings at a designated media center. Designated public spokespersons are the president of Duke Energy nuclear, the chief nuclear officer and his direct reports, and their designees.

Rev. 18-1 G-1 January, 2018

G.4.b Spokesperson Information Exchange State, county and licensee spokespersons/public information officers are co-located in the Joint Information Center (JIC) to promote a timely exchange and coordination of emergency information. If the JIC is not activated, or if the state or counties do not send public information officers to the JIC, information will be shared via a Joint Information System (JIS) utilizing email, common shared platforms such as W ebEOC, faxes or other co:rnmunication vehicles agreed upon by the affected agencies.

G.4.c Rumor Control A licensee liaison will work with state, county, and federal public information officers in the JIC or via

'the JIS to acknowledge rumors and determine the origin. A coordinated response will be made to deal with rumors or correct misinformation.

Customer inquiries are handled by our Customer Contact Centers. Employees are updated via the company intranet/portal. Elected officials and regulatory agencies are updated through our Corporate Communications and governmental affairs departments. Industry groups would assist in disseminating information to other industry groups.

G.5 News Media Training Sessions The licensee will annually provide the news media with information about emergency planning, radiation, and points of contact for release of public information in an emergency.

Rev. 18-1 G-2 January, 2018

H. Emergency Facilities and Equipment H.1 Technical Support Center (TSC)/Operations Support Center (OSC)

H. l.a Control Room. The Control Room is utilized for evaluation and control of the initial phase of an emergency, including corrective actions and notification and activation of McGuire, Duke Energy, state and local emergency response organizations. The Control Room has redundant (telephone and alternate) two-way communications with emergency centers and off-site agencies. See Figure F-1 for communication scheme.

H.l.b Technical Support Center. (Figure H-1) The Technical Support Center (TSC) is utilized for evaluation of plant status by knowledgeable plant, vendor, NRC and other support groups during an emergency. This center will also be utilized to direct the on-site and initial off-site aspects of an emergency. Anticipated occupants are defined in Emergency Planning Group Manual Section 1.1, On-site Emergency Organization.

The TSC has the following capabilities:

1. Redundant two-way communications with the Control Room, the OSC, the Emergency Operations Facility and the Nuclear Regulatory Commission Operations Center. See Figure F-2 for communication scheme.

2. Monitoring for direct radiation and airborne radioactive materials with local readout of radiation level and alarms if levels are exceeded.

3. Display, printout or trend record of comprehensive data necessary to monitor reactor system status and to evaluate plant system abnormalities, in-plant and off-site radiological parameters and meteorological parameters are available.

This capability is provided via the operator aid computer. Capabilities to access and display parameters, individually or in groups is provided.

4. Ready access to as-built plant drawings such as general arrangements, flow diagrams, electrical one-lines, instrument details, etc.

5. Radiological habitability during postulated radiological accidents to the same degree as the Control Room.

6. Provisions for staffing by the Station Manager (Emergency Coordinator),

advisors and representatives from the site as necessary. Room is also provided for NRC personnel. Space for up to 35 persons plus instrumentation displays are provided.

The TSC is located near the Control Room, on elevation 767, in the Service Building. The TSC is within one (1) minute walking distance from the Control Room. This is a permanent facility.

Rev. 18-1 H-1 January, 2018

H.1.c Operations Support Center. (Figure H-2) The Operations Support Center (OSC) is that place designated for Operations, Radiation Protection, Chemistry, Maintenance, IAE, and others as necessary, to report to in an emergency condition. This center will be used to brief and prepare site personnel for work assignments in support of the emergency condition. The OSC is located on the Auxiliary Building roof office, elevation 784'. Workspace and resources are shared with the Outage Control Center (OCC). The OSC shall have priority over the OCC if any emergency is declared during an outage. The OSC has adequate capacity and supplies including provisions for respiratory protection, protective clothing, portable lighting, portable radiation monitoring equipment and communications equipment.

H.1.d Alternate Facilities. (Figures H-9 and H-10) Alternate TSC and OSC facilities have been established in the McGuire Admin Building as a contingency. Communications equipment similar to that provided in the designated TSC and OSC facilities is available but not all regulatory required equipment/capability is provided.

H.2 Emergency Operations Facility (EOF)

The Emergency Operations Facility (EOF) is utilized for direction and control of all emergency and recovery activities with emphasis on the coordination of off-site activities such as communications with local, state and federal agencies, and coordination of corporate and other outside support. Anticipated occupants are the EOF organization and appropriate state and federal agency representatives.

The EOF has the following capabilities:

a. The capability for obtaining and displaying plant data and radiological information for each reactor at a nuclear power reactor site and for each nuclear power reactor site that the facility serves.

b. The capability to analyze plant technical information and provide technical briefings on event conditions and prognosis to licensee and offsite response organizations for each reactor at a nuclear power reactor site and for each nuclear power reactor site that the facility serves.

c. The capability to support response to events occurring simultaneously at more than one nuclear power reactor site if the emergency operations facility serves more than one site.

Rev. 18-1 H-2 January, 2018

The Common EOF in Charlotte serves as an alternate facility that would be accessible even if the site is under threat of or experiencing hostile action, to function as a staging area for augmentation of emergency response staff and having the following characteristics required collectively of the alternate facilities for use when onsite emergency facilities cannot be safely accessed during hostile action:

The capability for communication with the emergency operations facility, control room, and plant security.

The capability to perform offsite notifications.

The capability for engineering assessment activities, including damage control team planning and preparation.

The EOF has redundant two-way communications with the Technical Support Center and appropriate off-site support agencies. (See Section F).

The EOF is located at 526 South Church Street, Charlotte, NC in the Energy Center Phase II, third floor (Rooms 0300, 0330, 0331, 0332, 0333, 0334, 0335, 0336, 0337, 0337-A, 0340, 0341, 0342, 0343, 0343-A, 0343-B, 0344 and 0345). The EOF layout and location are shown on Figures H-3 thru H-5.

The Joint Information Center and Media Center are utilized for the origination of news briefings and interviews. Anticipated staffing includes the News Group personnel, industry and government representatives and support personnel. News media personnel can be accommodated for press conferences, etc., in the Media Center. (See Figure H-6 and H-7.)

The Joint Information Center has two-way communications with the Emergency Operations Facility and corporate headquarters.

The Joint Information Center (TIC) is located in Duke's Energy Center, 526 South Church Street, Charlotte, N.C. The JIC is located on the first floor, room ECI-0111.

The facilities and resources in the nc include:

Workspace

Telephones

Facsimile machines

Copy machines

Podium and PA system

Tone alert radio

TV monitor and VCR for real time viewing of the press conferences and taped review of news broadcasts from all three major networks

Status board

Wall charts dealing with nuclear site systems and evacuation zones

Name tags

Limited clerical support as needed

Meals during long term activation

Security escort to other TIC facilities as needed Rev. 18-1 H-3 January, 2018

The media center is located in Duke's Energy Center, 526 South Church Street, Charlotte, N.C.

The center is located on the first floor in the O.J. Miller Auditorium.

The facilities and resources in the Media Center include:

PA system and direct access to recording

18 telephones for news media

Court recorders for prompt press conference transcripts

Charts dealing with nuclear site systems and evacuation zones

Modem/computer connections for the news media

. Overhead projector

Slide projector

Screen

Press kits

News releases

Technical resources

Security, registration and badging H.3 State and Local Government Emergency Operations Centers See County and State Plans.

H.4 Activation and Staffing McGuire emergency response facilities (TSC, OC, EOF) are activated as required by the appropriate Emergency Response Procedure. Activation of the TSC, OSC, and EOF is required for Alert and higher emergency conditions. Timely activation and staffing of the Emergency Operations Facility is important to allow the Nuclear Station staff the ability to correct the situation with minimal interference from outside organizations. The Emergency Coordinator will perform the role and function of the EOF Director until activation of the EOF has taken place. The EOF Organization will be alerted and activated for Alert and higher emergency classifications.

H.5 Assessment Actions Onsite monitoring systems used to initiate emergency measures are defined in Section I. Those used for conducting assessment evaluations during any emergency condition are listed below:

H.5.a Meteorological. A description of the primary meteorological measurement facility is found in Appendix 2. These basic meteorological parameters are displayed in the Control Room, see Figure H-8, Generalized Meteorological System.

1. During periods of primary system unavailability, an alternate source of meteorological data is established as the NWS (NATIONAL WEATHER SERVICE) office. Wind direction and speed are from standard NWS instrumentation at conventional heights.

Wind direction from the NWS can replace the tower (60 m) wind direction.

Wind speed from the NWS can replace the lower tower (10 m) wind speed for Rev. 18-1 H-4 January, 2018

dose calculation purposes; it can also replace the tower (60 m) wind speed for transport speed considerations.

A monthly telephone contact, initiated by plant personnel, with the NWS office will be established to insure that this basic meteorological information can be accessed. See PT/O/A/4600/089.

2. The following field checks will be performed each week by plant personnel:

Wind Direction (a) Recorder Time Accuracy (b) Recorder Zero (c) Translator Zero (d) Translator Full Scale Wind Speed (a) Recorder Time Accuracy (b) Recorder Zero (c) Translator Zero (d) Translator Full Scale Delta - Temperature (a) Recorder Time Accuracy

3. Onsite meteorological instruments will be calibrated at a frequency specified by Selective Licensee Commitments. During calibration periods, basic meteorological data, characteristic of site conditions, will be accessible from the NWS. These instruments will be calibrated in accordance with approved procedures.

Hydrologic A hydrological description of the McGuire Nuclear Site is located in the MNS FSAR, Section 2.4.

Seismic A description of the seismic monitoring instrumentation and area seismology studies are found in McGuire FSAR, Sections 3.7 and 2.5 respectively.

Rev. 18-1 H-5 January, 2018

H.5.b Radiological Monitors Radiological monitors including process monitors, area monitors, post-accident monitoring equipment, effluent monitors, personnel monitoring devices, portable monitors and sampling equipment are described in various Radiation Protection procedures, the McGuire FSAR, Emergency Plan Implementing Procedures and Safety Evaluation Report.

H.5.c. Plant Parameters Equipment and instrumentation to monitor plant parameters such as reactor coolant pressure, temperature, levels, containment pressure, temperature, humidity, sump levels, hydrogen concentrations, system flow rates, status, line-ups, are included in operating and emergency procedures. Examples of specific instruments used for accident evaluation are given in Section I.

H.5.d Fire Detection Fire detection devices of the ionization-chamber and thermal type are located throughout the site.

H.6 Data. Monitoring Equipment and Analysis Facilities Provisions have been made and exist to obtain data from off-site agencies or monitoring equipment and analysis facilities. The provisions are described below:

a. Meteorological information is available from the National Weather Service as described in Section H.5.a. Monitoring of the Catawba River for hydrologic data is conducted within the Duke System of dams and hydro-electric facilities. Seismic data is available from the U.S. Geological Survey Office as provided for in the McGuire Procedure RP/O/A/5700/007 (Earthquake).

b. Radiological monitors for emergency environmental monitoring are provided in emergency kits. The established environmental monitoring network and sampling equipment in the surrounding area are also available to provide emergency assessment data. Environmental Radiological Monitoring equipment includes radioiodine and particulate continuous air samplers and thermoluminescent dosimeters. The thermoluminescent dosimeters are posted and collected in accordance with Table 1, Branch Technical Position, Rev. 1 of November, 1979. A procedure lists locations of posted thermoluminescent dosimeters and air samplers.

c. See Section C.3.

Rev. 18-1 H-6 January, 2018

H.7 Offsite Radiological Monitoring As described in H.6.b above.

H.8 Meteorology Instrumentation and Procedures See Section H.5.a.

H.9 Operations Support Center See Section H.1.c.

H.10 Emergency Equipment/Instrumentation Inspection, Inventory, Operational Check, Calibration McGuire Procedure PT/O/A/4600/088, Functional Check of Emergency Vehicle and Equipment, defines the inspection, inventory and operational checks required of emergency equipment. Various Radiation Protection procedures define the criteria for calibration of all monitoring equipment located in the emergency kits.

H.11 Emergency Kits Radiological Emergency kits are described m PT/O/A/4600/088, Functional Check of Emergency Vehicle and Equipment.

H.12 Receipt and Analysis of Field Monitoring Data Duke Energy's Emergency Operations Facility (Radiological Assessment Manager) will be the central point for the receipt of off-site monitoring data results and sample media analysis results collected by Duke personnel. Resources exist within the organization to evaluate the information and make recommendations based upon the evaluations. The Radiological Assessment Manager's group will perform these evaluations and make recommendations to the EOF Director for protective actions. The EOF Director is the individual responsible for making protective action recommendations to off-site agencies after activation of the EOF.

Rev. 18-1 H-7 January, 2018

McGuire Technical Suppo1t Center Layout Figure H-1 (3} W- MONITORS

//'"-*

u:::::****='!!;;;;8t:::*****=--****=----*:::2****,,c____J,,,:*;!L:===::2__---1/

Rev. 18-1 H-8 January, 2018

0 Briefing Room Mech, IAE1 Chem, RP Techs &

OPS co 2 - OPS Liaison 3 - Operations Supv. x-4475 4 - Asst OSC Manager. x-4952, 4518 5 - OSC Manager. x-4952 7 - Mnl Supv. x-4958 9- RP Tech 0 0 10-RPTech 11-RP Tech 12 - RP Supv. x-4978 13 - Log Keeper x-4292 14-Log Keeper#2 x-16-Nuclear Supply Chain x-4963 17 -Chemistry Supv. x-4960

<O Copier OUTSIDE BELL LINES 980-875-1952 980-875-1957 980-875-1958 0 Staffing Board OSC FAX 980-875-4423 RP Phones Supplies Radios Supplies pr;;:::s Figure H-2 Duke Energy McGuire Nuclear Site Operations Support Center O..._______.

Rev. 18-1 H-9 January, 2018

FIGURE H-3 DUKE ENERGY EMERGENCY RESPONSE MNS/CNS/ONS EOF GENERAL LOCATION r.'lcriulr$

'4ur.l'1i:JiP Sfntlcrr I .. /** ---- - ...., ......

I I ,,J.... **--

I /~****

1, M~O, l,'11 ..~*

- - - - -~':::~........, , /"

./

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1""1.****-7**::: ::::

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Alt~<,r, /

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/~1-.::ei.ur deu,il

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Rev. 18-1 H-10 January, 2018

The Media Center and Joint Information Center are in the Energy Center Phase I on the 1st floor.

The EOF is in the Energy Center Phase II on the 3rd floor.

Figure H-4 DUKE ENERGY GENERAL OFFICE RESPONSE FACILITY GENERAL OFFICE BUILDING LAYOUT- CHARLOTTE, NC S MINT ST "C

ENERGY .2:

CENTER DJ 526 S. Church St. .,..:

~ ci

-- I-C:

i:

~

s,W I

C CHURCH ST ....

....:I ti

~ ll. _J C:

~ ~

400 S. Tryon ns 3 :!E Duke Energy Center ~

TRYON ST Rev. 18-1 January, 2018 H-11

FIGUREH-5 Emergency Operations Facility EOFGENERALARRANGEMENT

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CONF Rl.4

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'E MOK[

HAFT 0 ~~ ~.".,;,,.

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Rev. 18-1 January, 2018 H-12

Figure H-6 Duke Energy

. Media Center I Stage I

Charts Podium

- - - - - - - - - - - - - - - - - - - (Rope)

(Media Seats) (Media Seats) 1191111118 11a11a11a 11911111111 11a11a11a 11111191111 11111191111 11a111111a lllllltlllll 1111111111a 11aaa11a Video Cameras Media Coordinator's State Tables Table (Partition)

Media Phones Court Reporters Projection Room Rev. 18-1 H-13 January, 2018

Figure H-7 Duke Energy Joint Information Center

~.:fl '

'., Ii tf=ttI tF STATEfCOUNTY

, ..... ~ f-1 F

/

Rev. 18-1 H-14 January, 2018

FIGUREH-8 McGuire and Catawba Nuclear Sites

,Generalized Met System Sensor (Met) and Instrumentation Computer U-1 _J Rad I U-2 Rad Room OAC I Data I OAC Data I TSC - Site LAN OAC SDS WORKSTATIONS Outputs to TSC, EOF, Control Room OSC display areas Read only mode for NRC and State State remote interrogation.

Rev. 18-1 H-15 January, 2018

FIGUREH-9 MCGUIRE NUCLEAR SITE ALTERNATE TECHNICAL SUPPORT CENTER (EXECUTIVE BOARD ROOM, ROOM 111, ADMIN. BUILDING)

Assistant NRC OPS OPS Elect NRC Resident Mgr Mgr.

Eng. Communicator PC 4951 4520 4962 875-4789(PMCL)

Mech. 875-1955 875-1953 4954 875-4770(HPN)

Engineer 875-4778 (ENS) 875-4788 (RSCL)

Reactor PC Engineer 4968 4959 PC Security RP Log Eng. Mgr.

Site Evacuation Manager Keep Coordinator

2211 Other TSC Position Locations

Site Evacuation Coordinator (EP Room 114) - *4458, *4977, *875-1951.

Offsite Communicator (EP Room 117 -- *4970, DEMNET, *Radio, *875-1951.

IT Support (CBX Equipment Room 112) -- *4248.

OAC Support (CBX Equipment Room 112) -- *4999.

Dose Assessor (SCR Room 100D) -- *4405.

Company Spokesperson (Rooms 118 and 141) -- *4400, *4419, *4233.

NRC (NRC Office, Room 126) -- *875-1681.

Other, use Jaguar Room as needed (Room 144) -- *4826.

Office Equipment

FAX (Mail Room, Room 116) -- *875-4506.

FAX (EP Room 114)-- *875-4382.

Copier (Mail Room, Room 116).

Copier (SA Room 170).

CBX (CBX Office in Admin. Building Lobby).

Indicates existing phones. All others are to be plugged in when the Alternate TSC is activated.

Rev. 18-1 H-16 January, 2018

FIGUREH-10 MCGUIRE NUCLEAR SITE ALTERNATE OPERATIONS SUPPORT CENTER (TRAINING ROOM TR155, ADMIN. BUILDING)

'i"'~*v O~cy*

9,'>'1, oe:,u5..* ~ ~ Mnt. Supv.

NLO NLO

~'$

~

"1" °'" Nuclear Supply Chain 4958 4974 4975 I

Ringdown Liaison 4963 Chair Storage Phone to TSC 4957 RP 4965 Mech 4973 RP OPS Supv.

4978 Briefing Outside 875-1958 PC Area 4967 Log Chem IAE 4956 4960 Keeper 4966 4972 OPS Chemistry Supv.

Supv.

Office Equipment

FAX, Mail Room, Room 116 -- *875-4506 .

FAX, EP, Room 114-- *875-4382 .

Copier, Mail Room, Room 116 .

Copier, SA, Room 170 .

CBX, CBX Office in Lobby.

Indicates existing phone. All others are to be plugged in when the Alternate OSC is activated.

Rev. 18-1 H-17 January, 2018

I. ACCIDENT ASSESSMENT To assure the adequacy of methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition.

1.1 Emergency Action Level Procedures Revision 6 ofNEI 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 , MNS conducted an EAL implementation upgrade project that produced the EALs discussed in Section D.

The EAL Technical Basis Document (Section D) and EAL Wallboards, will identify the system parameter and effluent parameter values which can be used to determine the emergency condition.

1.2 Onsite Capability and Resources to Provide Initial Values and Continuing Assessment 1.2.a. Post Accident Sampling The requirement to have reactor coolant and containment atmosphere post accident sample panels (PASS) has been deleted per NRC License Amendment 199/180 by letter dated September 17, 2001. As a result, Emergency Plan Implementing Procedures HP/1/B/1009/015, Unit 1 Nuclear Post-Accident Containment Air Sampling System Operating Procedure and HP/2/B/1009/015, Unit 2 Nuclear Post-Accident Containment Air Sampling System Operating Procedure have been deleted. These EPIPs were replaced by Radiation .

Protection group procedure HP/O/B/1009/032, Sampling Containment Atmosphere Under Accident Conditions. HP/O/B/1009/032 is W!! an EPIP or a part of the EPLAN. It is listed in this paragraph for reference purposes only. HP/O/B/1009/032 provides contingency methods for containment atmosphere sampling under accident conditions.

Also as a result of NRC License Amendment 199/180, OP/O/B/6200/090, PALSS Operation for Accident Sampling has been deleted from the EPLAN as an Emergency Plan Implementing Procedure. However, OP/O/B/6200/090 will be retained as a Chemistry group procedure to provide contingency methods for reactor coolant* and containment sump sampling under accident conditions. OP/O/B/6200/090 is not an EPIP or a part of the EPLAN. It is listed in this paragraph for reference purposes only.

1.2.b. Radiation and Effluent Monitors Radiological monitoring capabilities include process and effluent monitoring systems (FSAR 11.4); area monitoring system (FSAR 12.1.4); plus station portable monitoring instruments, laboratory counters and analyzers (FSAR 12.3.2.4), including emergency high-range instruments and air samplers.

Rev. 18-1 1-1 January, 2018

In addition, there are two (2) high range containment monitors, two (2) high range unit vent monitors, four (4) steam line monitors per unit and four (4) N-16 steam line monitors per unit.

1.2.c In-plant Iodine Instrumentation Radioiodine sampling cartridges are used for sampling containments and unit vents. Radiation Protection personnel are knowledgeable in the appropriate site procedures required and are trained in the equipment required to determine airborne iodine concentrations in the plant under all conditions. Procedures to determine airborne iodine concentrations will cover analyses to be done if counting room *capabilities are not available.

1.3.a/ Method For Determining Release Source Term 1.3.b Procedures HP/O/B/1009/006, HP/O/B/1009/010 and AD-EP-ALL-0202 are used on shift, in

. the TSC and/or EOF for the calculation of potential off-site doses based on a Design Basis Accident, release of primary coolant, or release of GAP activity situation scaled to actual containment monitor readings. Provisions for use of actual source terms exist in the procedures.

The magnitude of the release is based on actual effluent monitoring readings, plant system parameters (containment pressure), area meteorology and the duration of the release.

I.4 Effluent Monitor Readings Vs Onsite/Offsite Exposure The procedures referenced in l.3.a/1.3.b establish the relationship between effluent monitor readings and on-site/off-site exposures and contamination for various meteorological conditions.

1.5 Meteorological Information Availability Meteorological information will be available to the Emergency Operations Facility, the Technical Support Center, the Control Room through use of the Station Operator Aid Computer (OAC) and by direct telephone communication. Meteorological information will be available to the NRC through the Emergency Response Data System (ERDS), Health Physics Network (HPN) or by direct telephone communications with the individual responsible for making off-site dose assessments either at the Tec~cal Support Center or the Emergency Operations Facility. e '* - , **

Meteorological information will also be given to both the county Emergency Operations I Centers and the State of North Carolina during the follow-up notification ENF.

1.6 Release Rates/Projected Dose For Offscale Instrumentation If instrumentation used for dose assessment is offscale or inoperable, dose rates within the Reactor Building will be determined using procedure HP/O/B/1009/002, Alternative Method for Determining Dose Rate Within the Reactor Building, or HP/O/B/1009/006, Procedure for Quantifying High Level Radioactivity Release During Accident Conditions.

Rev. 18-1 I-2 January, 2018

1.7/ Field Monitoring Within E.P .z.

1.8 Field monitoring within the McGuire Emergency Planning Zone will be performed in I accordance with fleet and site specific procedures.

Two off-site field monitoring teams are comprised from site personnel and are under the direction of the Field Monitoring Coordinator. On-site monitoring is performed by Radiation Protection personnel under the direction of the OSC Radiation Protection Supervisor.

Procedures describe how to obtain the vehicles to be used, routes to be used, sampling and monitoring equipment to be used, locations of TLD's and directions for taking Kl tablets.

An emergency radio system is available for the field monitoring teams to use to relay information to the TSC/EOF. The state will be able to monitor the results of the field monitoring teams and relay results to the counties.

1.9 Detect and Measure Radioiodine Concentration in the EPZ Appropriate instrumentation to measure radioactivity in counts per minute (cpm) and determine dose rate in mrem/hr shall be used for detection and measurement of radioiodine concentration.

The air sample will be taken with a Portable Air Sampler equipped with a Silver Zeolite or equivalent cartridge and particulate filter. Air sampling results will be obtained through the use of a portable single channel Analyzer and appropriate gamma sensitive detector OR a count rate meter utilizing direct corrected count rate (ccpm) of Silver Zeolite or equivalent cartridge cross referenced against an estimated Iodine 131 µCi/cc (microcuries per cubic centimeter) concentration attachment.

Interference from the presence of noble gas and background radiation shall not decrease the minimum detectable activity of lE-7 µCi/cc (microcuries per cubic centimeter) under field conditions.

These samples taken by the offsite monitoring teams will be evaluated further by one of the available laboratory facilities described in Section C.3. A multi-channel analyzer will be used to perform this evaluation.

1.10 Relationship Between Contamination Levels and Integrated Dose/Dose Rates Provisions .for relating contamination levels, water, and air to dose rates for key isotopes is found in HP/O/B/1009/021.

1.11 Plume Tracking The state of North Carolina has arrangements to locate and track an airborne plume of radioactive materials. Duke Energy will have monitoring teams in the field, fixed TLD sites and the capability for airborne monitoring to assist in plume tracking.

Rev. 18-1 I-3 January, 2018

J. PROTECTIVE RESPONSE To assure that a range of protective actions is available for the plume exposure pathway for emergency workers and the public. Guidelines for protective actions during an emergency, consistent with Federal guidance, are developed and in place and protective actions for the ingestion exposure pathway appropriate to the locale have been developed.

To protect onsite personnel during hostile action and ensure the continued ability to safely shutdown the reactor and perform the functions of the emergency plan a range of protective actions are in place.

J.1. Onsite Alerting and Notification The means and time required to warn, alert and/or notify employees not having emergency assignments (non-essential), visitors, contractor and construction personnel and other individuals who may be on or passing through .the owner-controlled area are described in RP/O/A/5700/011, Conducting a Site Assembly, Site Evacuation or Containment Evacuation.

Methods to notify and alert onsite personnel (essential and non-essential) during hostile action activities are describe in AP/O/A/5500/047, Security Events and RP/O/A/5700/011, Conducting A Site Assembly, Site Evacuation or Containment Evacuation.

J.2 Evacuation Routes and Transportation The Operations Shift Manager/Emergency Coordinator or designee uses site and local area maps, information available from meteorological tower instrument readouts and current radiological data for determining the evacuation route. Evacuation routes for onsite individuals to suitable offsite locations, including alternatives for weather or radiological conditions is provided in RP/O/A/5700/011, Conducting a Site Assembly, Site Evacuation or Containment Evacuation.

J.3 Personnel Monitoring Radiation Protection emergency personnel survey teams equipped with portable monitoring instruments will monitor employees, visitors, contract workers and vehicles for contamination at the Relocation Sites. Monitoring will be performed in accordance with Radiation Protection procedure HP/O/B/1009/024, Personnel Monitoring for Emergency Conditions.

Rev. 18-1 J-1 January, 2018

J.4 Site Evacuation Procedures - Decontamination/Non-Essential Personnel Criteria Non-essential personnel may be evacuated from the plant site in the event of a Site Area Emergency and will be evacuated in the event of a General Emergency. Provisions are made for the decontamination of vehicles and personnel at an off-site location if the situation should warrant.

All members of the general public who are on-site must be evacuated if there is a possibility they may be exposed to dose rates in excess of any of the following:

External Radiation Level = 2 mrems/hr Airborne Radioactivity= 1 times DAC for an unrestricted area During hostile threat conditions that do not require Site Sheltering, Site Relocation of non-essential personnel to locations outside of the protected area are performed in accordance with AP/O/A/5500/047, Security Events and RP/O/A/5700/011, Conducting A Site Assembly, Site Evacuation or Containment Evacuation.

J.5 Personnel Accountability Within thirty minutes of a Site Assembly, all persons within the Protected Area at the McGuire Nuclear Site can be accounted for and any person(s) determined to be missing, will be identified by name. RP/O/A/5700/011, Conducting a Site Assembly, Site Evacuation or Containment Evacuation, provides for the accounting of personnel (on site) continuously thereafter.

When hostile threat conditions permit, personnel accountability is performed in accordance with RP/O/A/5700/011, Conducting A Site Assembly, Site Evacuation or Containment Evacuation.

J.6 Protective Equipment - Breathing Apparatus, Protective Clothes, KI Protective equipment and supplies will be distributed to respiratory qualified personnel remaining on site or arriving on site during the emergency to minimize the effects of radiological exposures or contamination. Protective measures will be utilized as follows:

Individual Respiratory Protection - Respiratory protective equipment will be used when airborne radioactivity levels exceed the appropriate limits specified in 10CFR20, Appendix B.

Self-contained breathing apparatus will also be used in areas that are deficient in oxygen or when fighting fires. Respiratory protective equipment will be issued by Radiation Protection or Safety and Health Services. Self-contained breathing apparatus are available with other fire fighting equipment for use by the site fire brigade.

Individual Thyroid Protection - All efforts should be made to utilize respiratory protective equipment to minimize ingestion and/or inhalation of radionuclides and to maintain internal Rev. 18-1 J-2 January, 2018

exposure below the limits specified in 10CFR20, Appendix B. However, if an unplanned incident involves the accidental or potential ingestion or inhalation of radioactive iodine, Potassium Iodide Tablets (Kl) are available to distribution by AD-EP-ALL-0204, Distribution of Potassium Iodide Tablets in the Event of a Radioiodine Release.

Use of Protective Clothing - Protective clothing will be issued when contamination levels exceed 1000 dpm/100 cm2 beta-gamma and 20 dpm/100 cm2 alpha of smearable contamination. Protective clothing items are located in the Change Rooms inside the Radiation Control Area, available for emergency use. Special fire-fighting protective clothing and equipment is available in designated site supply storage areas for use by fire brigade personnel.

J.7 Protective Actions Recommendations The Emergency Coordinator (Operations Shift Manager or Station Manager) or the EOF Director shall be responsible for contacting the state and/or local governments to give prompt notification for implementing protective measures within the plume exposure pathway, and beyond it if necessary.

Protective Action Guides are adopted from EPA 400-R-92-001 and in the State Plan guidance on the use of KI and are shown in Figure J-1. A flowchart to aid the Emergency Coordinator/EOF Director in making Protective Action Recommendations is also shown in Figure J-1. {PIP-G-03-606}

As described in section B.4, the Emergency Coordinator and the EOF Director are responsible for making protective action recommendations. Prior to activation of the EOF, the Emergency Coordinator will be responsible for making these recommendations. After activation of the EOF, the EOF Director assumes this responsibility. Protective action recommendations will be provided to the off-site authorities (states and counties) who are responsible for implementing public protective actions. The pre-established warning message format (Figure E-1) will be used in transmitting the recommendations.

The mechanism for making dose projections upon EOF activation is as follows:

The Radiological Assessment Manager is responsible for making dose projections on a periodic basis. Calculations are made using a computer based dose projection model to calculate projected dose to the population-at-risk for either potential or actual release conditions. For conditions in which a release has not occurred but fuel damage has taken place and radiation levels in the containment building atmosphere are significant, a scoping analysis will be performed to determine what recommendations would be made if containment integrity were lost at that time. The analysis will be based upon a design leak rate and upon a projected penetration failure indicated by a hole size of certain diameter. This analysis will include the use of actual containment pressure, realistic meteorology, and actual source term. A Total Effective Dose Equivalent and Committed Dose Equivalent thyroid dose will be calculated at various distances from the plant (site boundary, 2 miles, 5 miles, 10 miles and beyond if needed). These dose projections are compared to the Protective Action Guides in Figure J-1, which are derived from the "Manual of Protective Action Guides and Protective Actions for Rev. 18-1 J-3 January, 2018

Nuclear Incidents" (EPA-400-R-92-001) and in the State Plan guidance on use of.KI. Based on these comparisons, protective action recommendations are developed by the Radiological Assessment Manager. The Radiological Assessment Manager informs the EOF Director of the situation and recommendations for protective actions. {PIP-G-03-606}

If dose projections show that PAGs have been exceeded at 10 miles, the dose assessment code and in-field measurements, when available, shall be used to calculate doses at various distances down wind to determine how far from the site PAG levels are exceeded. The Radiological Assessment Manager shall forward the results to the EOF Director who will communicate this information to the offsite authorities.

J.8 Evacuation Time Estimates An Analysis of Evacuation Time Estimates is available at the site and a summary of the Time Estimates is included in Figure J-3 and Appendix 4.

Under normal weather and for the critical time period (weekday during school hours), the maximum evacuation time for the McGuire EPZ is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 35 minutes. The critical component in the evacuation is the permanent resident population, all other segments of the population can be evacuated in less than the maximum time.

Under adverse weather conditions (winter storm), the evacuation time for the McGuire EPZ is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 40 minutes. This evacuation time assumes evacuation of the entire EPZ. Figure J-3 provides more detailed information including evacuation times for individual zones. Appendix 4 discusses the ETE used by the site, state and local planners.

A description of the methods and assumptions used in developing the analysis of evacuation time estimates is included in the current Evacuation Time Estimate study for McGuire Nuclear Site. (MNS-ETE-12132012, Rev. 000; MNS EVACUATION TIME ESTIMATES (ETE)

DATED December 2012). The Evacuation Time Estimates will be considered in evaluating protective action recommendations from the Technical Support Center or the Emergency Operations Facility. A copy of the most recent study is available in the MNS Master File under MNS-ETE-12132012.000 or EP Office area.

An updated ETE analysis will be submitted to the NRC under §50.4 no later than 365 days after MNS determination that the criteria for updating the ETE have been met and at least 180 days before using it to form protective action recommendations and providing it to State and local governmental authorities for use in developing offsite protective action strategies.

Rev. 18-1 J-4 January, 2018

The criteria for determination that an updated ETE analysis have been met:

a) The availability of the most recent decennial census data from the U.S. Census Bureau:

OR b) If at any time during the decennial period, the EPZ permanent resident population increases such that it causes the longest ETE value for the 2-mile zone or 5-mile zone, including all affected Emergency Response Planning Areas, or for the entire 10-mile EPZ to increase by 25 percent or 30 minutes, whichever is less, from the currently NRC approved or updated ETE.

During the years between decennial period censuses MNS will estimate EPZ permanent resident population changes once a year, but no later than 365 days from the date of the previous estimate, using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. MNS will maintain these estimates so that they are available for NRC inspection during the period between decennial censuses and shall submit these estimates to the NRC with any updated ETE analysis.

MNS ETE analysis, using the 2010 decennial census data from the U.S. Census Bureau, was submitted to the NRC via §50.4 on December 13, 2012.

J.9 Implementing Protective Measures If protective actions for any off-site location are deemed necessary, the emergency management agency of the affected County, in conjunction with the appropriate State agency (NC-Department of Public Safety) has the legal authority and responsibility for initiating protective measures for the general public in the plume exposure pathway EPZ including evacuation of these areas. Use of sheltering as an alternative to evacuation for impediments to evacuation and special populations is a decision that will be made by the offsite officials.

Sheltering in lieu of evacuation should also be considered during a short term release. A short term release is any release that can be accurately projected to be less than the affected protective action zone's evacuation time. An example would be a "puff release". In addition, sheltering may be appropriate (when available) for areas not designated for immediate evacuation because: 1) it positions the public to receive additional instructions; and 2) it may provide protection equal to or greater than evacuation. {PIP G-04-337} Public notification of the emergency, the resources used to determine if an evacuation is necessary, the evacuation routes, and the methods used for evacuating persons in the plume exposure pathway EPZ are outlined in the appropriate County and State emergency plans.

See County and State Plans for more detailed information.

For hostile action events, a range of protective actions for onsite workers including site relocation of essential personnel from potential target buildings, timely evacuation of non-essential site personnel, dispersal of critical personnel to safe locations, on site sheltering of Rev. 18-1 J-5 January, 2018

personnel and accountability of personnel after the attack are provided in emergency plan implementing procedures AP/O/A/5500/047, Security Events and RP/O/A/5700/011, Conducting A Site Assembly, Site Evacuation or Containment Evacuation.

J.10 Implementation of Protective Measures for Plume Exposure Pathway J.10.a EPZ Maps Figures J-1 and 2 describe the EPZ's, government jurisdictions and evacuation zones for McGuire Nuclear Site. Evacuation routes are displayed in Figure J-5.

J.10.b EPZ - Population Distribution Map Figure J-6 describes the population distribution by Emergency Planning subzone. The FSAR describes the population distribution by sector.

J.10.c EPZ - Population Alerting and Notification As described in Appendix 3 of this plan, a system exists for alerting and notifying the population (resident and transient) within the EPZ areas. This system is activated by the county and state organization and includes the use of large fixed-site sirens and the Emergency Alert System (EAS). A back-up means of alerting and notification is described in the State and County Emergency Plans.

J.10.d EPZ - Protecting Immobile Persons See County and State Plans.

J.10.e Use ofRadioprotective Drugs For Persons in EPZ See County and State Plans.

J.10.f Conditions For Use ofRadioprotective Drugs See County and State Plans.

J.10.g State/County Relocation Plans See County and State Plans.

J.10.h Reception Center Locations See County and State Plans.

J.10.i Evacuation Route - Traffic Capacities Rev. 18-1 J-6 January, 2018

See County and State Plans.

J. l O.j Evacuated Area Access Control See County and State Plans.

J.10.k Planning For Contingencies in Evacuation See County and State Plans.

J.10.1 State/County Evacuation Time Estimates The estimates shown in Appendix 4 are references in the County and State Pla11s.

J.10.m Bases For Protective Action Recommendations Figure J-1 describes the considerations used by Duke management in developing protective action recommendations.

J .11 Ingestion Pathway Planning See County and State Plans.

J.12 Reception Center- Registering & Monitoring See County and State Plans Rev. 18-1 J-7 January, 2018

Figure J-1 1 of3 Guidance for Offsite Protective Actions INITIAL PAR Determination GE Has Been Note 1: Protective Action Zones (P AZs) are defined in Table 1 on page 3. SIP is Shelter in Place.

Declared Note 2: Offsite Agencies have provided prior knowledge of offsite impediments to evacuation (such as flooding, bridge/road closures, adverse weather, traffic control not in place, etc.) AND specifically requested that site NOT issue an evacuation PAR.

Note 3: RPSA criteria: Only Applicable to First PAR after GE Declaration AND two of the following exist from EAL Wall Chart (IF RPSA conditions can NOT be confirmed, THEN answer NO)

1. LOSS of Containment Barrier per Table Fl (Fission Product Barrier Threshold Matrix)

AND

2. POTENTIAL LOSS of Containment Barrier per Containment Radiation Monitor EMF51A or 5 lB per Table F2 (Containment Radiation R/hr EMF 51A & B).

(See Note 1) OR

Evacuate 2-mile Radius There is a significant radiological release projected to exceed RG 1 EAL at the site boundary within an hour. ,

Evacuate 2-5 miles Downwind

SIP 5-10 miles Downwind (See Note 1)

Evacuate 2-mile Radius No

Evacuate 2-5 miles Downwind (See Note 1)

SIP 2-mile Radius

SIP 2-5 miles Downwind Recommend the consideration of Recommend the KI use by the public. consideration of KI use by the public GO TO Page 2 for EXPANDED PAR Rev. 18-1 J-8 January, 2018

FigureJ-1 2 of3 EXPANDED PAR Determination Note 1: Protective Action Zones (P AZs) are defined in Table 1 on page 3. SIP is Shelter in Place.

(IF a PAZ has been accurately selected for evacuation, it shall remain selected.) From INITIAL PAR Note 2: Offsite Agencies have provided prior knowledge of offsite impediments to evacuation (such as flooding, bridge/road closures, adverse weather, traffic control not in place, etc.) AND specifically requested that site NOT issue an evacuation PAR.

Note 3: A short-term release is one that can be accurately projected to be< three hours and controlled Continuous Assessment by the licensee. This consideration would typically apply to controlled venting of containment. Evaluate PAR based on changes in any of the following:

Increase in dose assessment projected values.

Note 4: Plant conditions exist which would require the classification of a General Emergency per the

Increase in field team measured values.

EALs. This does NOT include consideration of offsite dose-based EALs.

Shift in 15-minute average wind direction resulting in additional sectors being affected. (Table 3 on page 3)

Offsite Agencies provides information that offsite impediments no longer exist.

Hostile action based event has been terminated.

IF RPSA, when safer to do so consider evacuation of SIP P AZs based upon radiological assessment and discussions with Offsite Agencies.

No Yes (See Note 1)

Evacuate PAZ(s) that exceed 1jf;,;

(See Note 1) PAGs. If projected dose exceeds /,.A,'t---,:lt

Evacuate 2-mile Radius 5-Rem Thyroid CDE then ,;lR;

SIP 2-mile Radius

Evacuate 2-5 miles Downwind recommend the consideration of ..¢Ji

SIP 2-5 miles Downwind >{}',

Kl use by the public. ;<,,,~,I

SIP additional PAZ(s) that exceed PAGs Rev. 18-1 J-9 January, 2018

FIGURE J-1 3 OF3 GUIDANCE FOR OFFSITE PROTECTIVE ACTIONS Table 1 Protective Action Zones Wind Direction 2Mile 2-5 Miles 5-10 Miles Radius Downwind Downwind 0.1 - 22.5 B,C,L,M D,O,R E,F,S 22.6-45.0 B,C,L,M D,O,R E,Q,S 45.1 -67.5 B,C,L,M D,N,O,R E,P,Q,S 67.6-90.0 B,C,L,M D,N,O,R P,Q,S 90.1 -112.5 B,C,L,M N,O,R K,P,Q,S 112.6 - 135.0 B,C,L,M A,N,O,R I,K,P,Q,S

135.1 - 157.5 B,C,L,M A,N,O l,K,P,Q 157.6-180.0 B,C,L,M A,N H,I,J,K,P 180.1 -202.5 B,C,L,M A,N G,H,l,J,K,P 202.6 - 225.0 B,C,L,M A,D,N G,H,l,J,K,P 225.1 - 247.5 B,C,L,M A,D F,G,H,l,J 247.6- 270.0 B,C,L,M A,D F,G,H,I,J 270.1-292.5 B,C,L,M A,D E,F,G,H,J 292.6- 315.0 B,C,L,M A,D,R E,F,G 315.1- 337.5 B,C,L,M D,R E,F,G,S 337.6 - 360.0 B,C,L,M D,R,O E,F,S Table 2 PROTECTIVE ACTION GUIDES (PAGs)

(Projected Dose or Field Measurements) Total Effective Dose Committed Dose Equivalent (TEDE) Equivalent (CDE) Thyroid

               > 1 Rem                  >5Rem Table 3 WIND SPEED/DIRECTION ENF                          Line 9 Radiation Protection Manager McGuire SDS             Group Display ERO RDS DPC Meteorological Lab              704-382-0139 704-373-7896 National Weather Service            864-879-1085 Greer, S.C                 800-268-7785 Rev. 18-1 J-10                              January, 2018

Figure J-2

                                                           . f D escnp,10n  o fE vacuaf10n R egions Degrees From                                                    Sub-Zone Region     Description        North:    A    B      C     D     E     F    G     H     I     J    K L M N 0 p Q    R     s ROI       2-MileRing           NIA          X      X                                               X X R02       5-Mile Ring          NIA     X    X      X     X                                         X X X X        X R03         FullEPZ            NIA     X    X      X     X     X     X    X     X     X    X     X X X X X X X    X     X R04     Catawba County         NIA                                                               X ROS      Gaston County         NIA                                                                                X     X R06      Iredell County        NIA     X                                              X    X R07     Lincoln County         NIA                                                                 X X X X X X R08       Mecklenburg          NIA                       X X    X      X           X     X    X     X County Evacuate 2-Mile Radius and Downwind to 5 Miles Wind Direction   Degrees From                                                    Sub-Zone Ree:ion       From:           North:    A    B      C     D     E     F    G     H     I     J    K L M N 0 p Q    R     s R09         N,NNE           0.1 - 45.0      X      X     X                                         X X            X RIO         NE,ENE         45.1 - 90.0      X      X                                               X X   X        X Rll        E,ESE, SE      90.1 - 157.6      X      X                                               X X X X R12           SSE        157.5 - 180.0      X      X                                               X X X R13             s        180.1 - 202.5 X    X      X                                               X X X R14        SSW,SW        202.6 - 247.5 X    X      X                                               X X R15         WSW,W        247.6 - 292.5 X    X      X     X                                         X X R16    WNW,NW,NNW        292.6 - 360.0      X      X     X                                         X X Evacuate 5-Mile Radius and Downwind to the EPZ Boundary Wind Direction   Degrees From                                                    Sub-Zone Ree:ion       From:           North:    A    B      C     D     E     F    G     H     I    J     K L M N 0 p Q    R     s R17             N           0.1 - 22.5 X    X      X     X     X     X                             X X X X        X     X R18           NNE          22.6 - 45.0 X    X      X     X     X                                   X X X X        X     X R19            NE          45.1 - 67.5 X    X     X      X     X                                   X X X X   X    X     X R20           ENE          67.6- 90.0  X    X      X     X                                         X X X X   X    X     X R21             E         90.1 - 112.5 X    X     X      X                                         X X X X X X    X     X R22           ESE        112.6 -135.0  X    X     X      X                                         X X X X X X    X R23            SE        135.1 - 157.6 X    X     X      X                                       X X X X X X X    X R24           SSE        157.5 - 180.0 X    X     X      X                            X          X X X X X X      X R25             s        180.1 - 202.5 X    X     X      X                            X    X     X X X X X X      X R26           SSW        202.6 - 225.0 X    X     X      X                      X     X    X     X X X X X X      X R27           SW         225.1 - 247.5 X    X     X      X                X     X     X    X       X X X X        X R28          WSW         247.6 - 270.0 X    X     X      X           X    X     X          X       X X X X        X R29            w         270.1 - 292.5 X    X     X      X           X    X     X                  X X X X        X J-11                                  Rev. 18-1 January, 2018

Figure J-2 Degrees From Sub-Zone Region Description North: A B C D E F G H I J K L M N 0 p Q R s R30 WNW 292.6 - 315.5 X X X X X X X X X X X R31 NW 315.1 - 337.5 X X X X X X X X X X X X R32 NNW 337.6 - 360.0 X X X X X X X X X X X Staged Evacuation Mile Radius Evacuates, then Evactiate Downwind to 5 Miles Wind Direction Degrees From Sub-Zone Re1!ion From: North: A B C D E F G H I J K L M N 0 p Q R s R33 N,NNE 0.1 - 45.0 X X X X X X R34 NE,ENE 45.1 - 90.0 X X X X X X R35 E,ESE, SE 90.1 - 157.6 X X X X X X R36 SSE 157.5 - 180.0 X X X X X R37 s 180.1 - 202.5 X X X X X X R38 SSW,SW 202.6 - 247.5 X X X X X R39 wsw,w 247.6 - 292.5 X X X X X X R40 WNW,NW,NNW 292.6 - 360.0 X X X X X R41 5-Mile Ring NIA X X X X X X X X X Sub-Zone(s) Shelter-in-Place Sub-Zone(s) Shelter-in-Place Sub-Zone(s) Evacuate until 90% ETE for ROI then Evacuate J-12 Rev. 18-1 January, 2018

Figure J-3 MNS ETE Based on 2010 Census Time to Clear the Indicated Area of 90 Percent of the Affected Population Summer Summer Summer Winter Winter Winter Winter Summer Midweek Midweek Midweek Weekend Midweek Weekend Weekend Midweek (1) (2) (3) (4)

                                                  .. Weekend (5)          (6)      (7)     . (8)

(9) (10) '" (11) Weekend (12) (13) (14.) Scenario: Midday Midday Evening Midday Midday Evening Midday Midday Good Good Good Good Good Good Special Roadway Rain Rain Rain Ice Rain Ice Weather Weather Weather Weather Weather Weather Event Impact Entire 2-Mile Region, 5-Mile Region, EPZ and each County ROI 2:35 2:50 1:50 1:50 1:50 2:45 2:55 3:10 1:50 1:50 1:55 1:50 1:50 2:40 (B,C,L,M) R02 2:50 3:10 2:25 2:45 2:20 2:45 3:10 3:35 2:25 2:40 3:00 2:20 2:25 3:30 (A,B,C,D,L,M,N,O,R) R03 4:35 5:00 3:30 3:55 3:10 4:35 5:00 5:40 3:30 3:50 4:10 3:10 3:40 4:50 (All Sub-Zones) R04 2:25 2:25 2:00 2:00 2:00 2:25 2:25 2:25 2:00 2:00 2:00 2:00 2:00 2:25 (K) ROS 2:20 2:30 2:10 2:15 2:10 2:20 2:30 2:35 2:05 2:15 2:25 2:05 2:10 2:20

        <R,S)

R06 2:40 2:50 2:25 2:35 2:20 2:40 2:55 3:15 2:25 2:35 2:55 2:20 2:25 3:25 (A,l,J) R07 3:30 3:40 3:05 3:20 2:50 3:20 3:30 3:55 2:50 3:05 3:25 2:40 3:05 3:30

  <L,M,N,O,P ,Q)

ROS 4:35 5:00 3:30 3:55 3:10 4:40 5:05 5:45 3:30 3:50 4:10 3:10 3:35 4:50 (A,B,C,D,E F G.m 2-Mile Re~ ion and KeJ hole to 5 Miles R09 2:35 2:50 2:20 2:35 2:30 2:35 2:50 3:10 2:20 2:30 2:55 2:30 2:25 3:35 IB,C,D,L,M,R) RIO 2:30 2:40 1:55 1:55 1:55 2:35 2:45 2:55 1:55 1:55 2:00 1:55 1:55 2:30 (B,C,L,M,O) Rll . 2:50 2:30 2:35 1:55 1:55 1:55 2:30 2:35 1:55 1:55 2:00 1:55 1:55 2:30 (B,C,L,M,N,O) RI2 2:30 2:40 1:55 1:55 1:55 2:35 2:40 2:55 1:55 1:55 1:55 1:55 1:55 2:30 IB,C,L,M,N) R13 2:20 2:40 2:10 2:20 2:00 2:20 2:40 3:00 2:10 2:20 2:40 2:00 2:10 3:45 (A,B,C,L,M,N) RI4 2:25 2:40 2:10 2:20 2:00 2:25 2:45 3:00 2:10 2:20 2:35 2:00 2:10 3:40 (A,B,C,L,M) J-13 Rev. 18-1 January, 2018

Figure J-3 MNS ETE Based on 2010 Census Summer Summer Summer Winter Winter Winter Winter Summer Midweek Midweek Midweek Weekend Midweek Weekend Weekend Midweek Weekend - - Weekend

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

Scenario: (]) (2) (3) (4) (5) (6) (7) (8) (9) (I 0) (11) (12) (13) (14) Midday Midday Evening: Midday Midday Evening: Midday Midday Good Good Good Good Good Good Special Roadway Rain Rain Rain Ice Rain Ice Weather Weather Weather Weather Weather Weather Event Impact 2-Mile Rei ion, and Ke 11hole to 5 Miles R15 2:50 3:15 2:30 2:45 2:25 2:50 3:10 3:35 2:25 2:40 3:05 2:20 2:30 3:35 (A,B,C,D,L,M) R16 2:35 2:50 2:25 2:35 2:30 2:35 2:50 3:10 2:20 2:30 2:55 2:30 2:25 3:40 IB.C,D,L,M) 5-Mile Reg:ion and Keyhole to EPZ Boundary R17 (A,B,C,D,E,F,L,M,N, 4:10 4:40 3:15 3:35 3:00 4:20 4:40 5:15 3:10 3:30 3:50 2:55 3:10 4:25 O,R.S) R18 (A,B,C,D,E,L,M,N, 2:45 3:05 2:30 2:45 2:30 2:50 3:05 3:25 2:30 2:40 3:00 2:25 2:30 3:15 O,R,S) R19 (A,B,C,D,E,L,M,N, 2:55 3:15 2:35 2:50 2:35 2:55 3:15 3:35 2:35 2:45 3:05 2:35 2:35 3:20 O.Q,R,S) R20 (A,B,C,D,L,M,N, 2:45 3:10 2:25 2:40 2:20 2:45 3:00 3:30 2:20 2:35 2:55 2:20 2:25 3:25 O,Q,R,S) R21 (A,B,C,D,L,M,N, 3:00 3:15 2:35 2:55 2:30 2:55 3:15 3:40 2:30. 2:45 3:10 2:30 2:35 3:30 O,P,Q,R,S) R22 (A,B,C,D,L,M,N, 3:00 3:20 2:35 2:55 2:35 3:00 3:15 3:40 2:30 2:45 3:10 2:30 2:40 3:30 O,P,Q,R,S) R23 (A,B,C,D,K,L,M,N, 3:05 3:20 2:40 2:55 2:35 3:00 3:15 3:45 2:30 2:45 3:10 2:30 2:40 3:35 O,P,Q,R) R24 (A,B,C,D,I,K,L,M,N, 3:05 3:25 2:45 3:00 2:45 3:05 3:20 3:45 2:35 2:50 3:15 2:40 2:45 3:30 O,P,R) J-14 Rev. 18-1 January, 2018

Figure J-3 MNS ETE Based on 2010 Census

                                                                                        ~ '.

Summer Summer Summer Winter Winter Winter Winter Summer Midweek Midweek Midweek Weekend Midweek Weekend Weekend Midweek Weekend Weekend Scenario: (1) (2) (3) (4) (5) (6) (7) (8) (9) (10) (11) (12) (13) (14) Midday Midday Evenin2 Midday Midday Evenin2 Midday Middav Good Good Good Good Good Good Special Roadway Rain Rain Rain Ice Rain Ice Weather Weather Weather Weather Weather Weather Event Impact 5-Mile Re2ion and Keyhole to EPZ Boundary R25 (A,B,C,D,I,J,K,L,M, 3:00 3:20 2:45 3:00 2:40 3:05 3:20 3:45 2:40 2:55 3:15 2:35 2:45 3:30 N,O,P,R) R26 (A,B,C,D,H,I,J,K,L, 3:15 3:35 2:50 3:05 2:40 3:15 3:30 4:00 2:45 3:00 3:25 2:35 2:50 3:45 M,N,O,P,R) R27 (A,B,C,D,G,H,I,J, 3:45 4:05 3:10 3:30 2:55 3:40 4:05 4:35 3:10 3:25 3:50 2:50 3:15 4:35 L,M,N,0,R) R28 (A,B,C,D,F,G,H,J, 4:30 5:00 3:30 3:50 3:05 4:35 5:05 5:35 3:25 3:45 4:10 3:00 3:35 4:50 LM,N,O,R) R29 (A,B,C,D,F,G,H,L, 4:30 4:55 3:25 3:45 3:00 4:30 4:55 5:35 3:20 3:35 4:05 3:00 3:25 4:40 M,N,0,R) R30 (A,B,C,D,F ,G,L,M, 4:25 4:45 3:15 3:40 3:00 4:25 4:50 5:20 3:15 3:35 4:00 3:00 3:25 4:35 N,O,R) R31 (A,B,C,D,E,F ,G,L, 4:35 4:55 3:35 3:45 3:05 4:30 5:05 5:30 3:25 3:45 4:05 3:05 3:35 4:40 M,N,O,R) R32 (A,B,C,D,E,F,L,M, 4:10 4:35 3:10 3:30 2:55 4:20 4:40 5:10 3:i5 3:30 3:50 2:55 3:15 4:20 N,O,R) J-15 Rev. 18-1 January, 2018

Figure J-3 Summer Summer Winter Winter Winter Winter Winter Midweek Midweek Good Good Ice Ice Weather Weather Weather Weather Weather Weather Sta ed Evacuation - 2 Mile Re ion and Ke hole to 5 miles R33 3:10 3:20 3:10 3:20 3:30 3:10 3:20 3:30 3:15 3:20 3:30 3:30 3:15 4:00

 ,C,D,L,M,R R34 2:30  2:40    2:25  2:25     2:25         2:35      2:45  2:55       2:25      2:25 2:25   2:25    2:25        2:30 (B,C,L,M,O)

R35 2:40 2:45 2:35 2:40 2:35 2:40 2:50 2:55 2:35 2:40 2:45 2:35 2:35 2:40

 ,C,L,M,N,0)

R36 2:40 2:45 2:35 2:35 2:35 2:40 2:45 2:55 2:35 2:35 2:40 2:35 2:35 2:40 2:45 2:55 2:45 2:50 2:50 2:45 2:55 3:05 2:45 2:55 3:00 2:55 2:45 3:35 2:50 2:55 2:50 2:55 2:55 2:50 2:55 3:10 2:50 2:55 3:05 2:55 2:45 3:45 A,B,C,L,M R39 3:15 3:25 3:15 3:20 3:25 3:15 3:30 3:45 3:15 3:20 3:35 3:25 3:15 4:00 A,B,C,D,L,M R40 3:15 3:20 3:15 3:20 3:30 3:15 3:20 3:35 3:15 3:20 3:30 3:30 3:15 4:00 B,C,D,L,M) R41 (A,B,C,D,L,M, 3:15 3:25 3:10 3:15 3:20 3:15 3:20 3:40 3:10 3:20 3:30 3:20 3:10 3:55 N,0,R) J-16 Rev. 18-1 January, 2018

Figure J-4 MNS ETE Based on 2010 Census Time to Clear the Indicated Area of 100 Percent of the Affected Population Summer Summer Summer Winter Winter Winter Winter Summer Midweek Midweek Midweek Weekend Midweek Weekend Weekend Midweek Weekend Weekend Scenario: (1) ' (2) (3) (4) (5) (6) (7) (8) (9) (10) (11) (12) (13) (14) Midday Midday Evening Midday Midday Evening Midday Midday Good Good Good Good Good Good Special Roadway Rain Rain Rain Ice Rain Ice Weather Weather Weather Weather Weather Weather Event Impact Entire 2-Mile Region, 5-Mile Region, EPZ and each County ROl 4:30 4:30 4:30 4:30 4:30 4:30 4:35 4:35 4:30 4:30 4:30 4:30 4:30 4:30 lB,C,L,M) R02 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:00 4:35 4:35 4:35 4:35 4:35 4:45 (A.B.C D L.M N O R) R03 6:10 6:50 5:00 5:25 5:05 6:20 7:15 7:55 5:05 5:15 5:40 5:05 5:00 7:10 (All Sub-Zones) R04 {K) 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 ROS 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 (R,S) R06 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:20 (A,l,J) R07 4:35 4:50 4:35 4:35 4:35 4:35 4:35 5:05 4:35 4:35 4:35 4:35 4:35 4:35 (L.M.N,O.P.rn ROS 6:05 6:50 4:45 5:10 4:40 6:20 7:00 7:45 4:40 5:15 5:40 4:40 4:50 7:05 (A,B,C,D,E,F,G,ID 2-Mile Re~ ion and Keyhole to 5 Miles R09 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:15 (B,C,D,L,M,R) RlO 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 lB,C,L,M,0) Rll 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 (B,C,L,M,N,O) R12 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 ffi,C,L,M,N) J-17 Rev. 18-1 January, 2018

Figure J-4 Summer Summer .Summer Winter Winter Winter Winter Summer Midweek Midweek Midweek Weekend Midweek Weekend Weekend Midweek Weekend Weekend Scenario: (]) (2) (3) (4) (5) (6) (7) (8) (9) (I 0) ( 11) (12) (13) (14) Midday Midday Evening Midday Midday Evening Midday Midday Good Good Good Good Good Good Special Roadway Rain Rain Rain Ice Rain Ice Weather Weather Weather Weather Weather Weather Event Impact Entire 2-Mile Region, and Keyhole to 5 miles R13 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:30 (A,B,C,L,M,N) R14 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:30 (A,B,C,L,M) R15 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:45 (A,B,C,D,L,M) R16 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:40 4:35 4:35 4:35 4:35 4:35 5:20 (B,C,D,L,M) 5-Mile Region and Keyhole to EPZ Boundary R17 (A,B,C,D,E,F,L,M,N, 5:55 6:10 5:00 5:10 5:00 6:15 6:30 7:20 5:05 5:10 5:35 5:00 5:05 6:40 O,R,S) R18 (A,B,C,D,E,L,M,N, 4:40 4:40 4:40 4:40 4:40 4:40 4:40 5:10 4:40 4:40 *4:40 4:40 4:40 5:40 O,R,S) R19 (A,B,C,D,E,L,M,N, 5:15 6:00 4:55 5:10 4:55 5:35 6:05 6:20 4:55 5:05 5:30 4:55 5:00 5:40 O,Q,R,S) R20 (A,B,C,D,L,M,N, 4:40 4:40 4:40 4:40 4:40 4:40 4:40 5:05 4:40 4:40 4:40 4:40 4:40 5:45 O,Q,R,S) R21 (A,B,C,D,L,M,N, 4:40 5:05 4:40 4:40 4:40 4:40 5:05 5:15 4:40 4:40 4:40 4:40 4:40 5:45 O,P,Q,R,S) R22 (A,B,C,D,L,M,N, 4:40 4:45 4:40 4:40 4:40 4:40 4:40 5:10 4:40 4:40 4:40 4:40 4:40 5:40 O,P,Q,R) J-18 Rev. 18-1 January, 2018

Figure J-4 Summer Summer Summer Winter Winter Winter Midweek Midweek Midweek Good Good Ice Ice Weather Weather Weather Weather Weather Weather 5-Mile Re ion and Ke hole to EPZ Bounda R23 (A,B,C,D,K,L,M,N, 4:40 5:10 4:40 4:40 4:40 4:40 4:50 5:25 4:40 4:40 4:45 4:40 4:40 5:45 OP R R24 (A,B,C,D,I,K,L,M,N, 4:40 5:05 4:40 4:40 4:40 4:40 4:50 5:20 4:40 4:40 4:40 4:40 4:40 5:40 O,P,R R25 (A,B,C,D,I,J,K,L,M, 4:40 5:05 4:40 4:40 4:40 4:40 4:45 5:20 4:40 4:40 4:40 4:40 4:40 5:40 N,O,P,R R26 (A,B,C,D,H,I,J,K,L, 4:40 5:05 4:40 4:40 4:40 4:40 4:45 5:20 4:40 4:40 4:40 4:40 4:40 6:05 MNOPR R27 (A,B,C,D,G,H,I,J, 5:05 5:25 4:40 4:40 4:40 4:55 5:30 6:05 4:40 4:40 4:40 4:40 4:40 6:40 L,M,N,O,R R28 (A,B,C,D,F,G,H,J, 6:00 6:30 4:40 4:55 4:40 6:15 7:05 7:30 4:40 4:45 5:30 4:40 4:55 7:10 L,M,N,O,R) R29 (A,B,C,D,F,G,H,L, 6:10 6:25 4:40 5:00 4:40 6:20 7:00 7:30 4:40 4:45 5:25 4:40 4:40 7:05 M,NOR R30 (A,B,C,D,F,G,L,M, 6:10 6:35 4:40 4:45 4:40 6:15 6:55. 7:20 4:40 4:45 5:20 4:40 4:40 7:00 N,O,R R31 (A,B,C,D,E,F,G,L, 6:05 6:45 4:45 5:10 4:40 6:15 7:15 7:50 4:50 5:05 5:35 4:40 4:50 6:55 M,N,O,R) R32 6:00 6:30 4:40 4:50 4:40 6:05 6:35 7:00 4:55 5:00 5:20 4:40 4:55 6:30 (A,B,C,D,E,F ,L,M, N,O,R J-19 Rev. 18-1 January, 2018

Figure J-4 Summer Summer Summer Winter Winter Winter Winter Winter Midweek Midweek Good Good Good Ice Weather Weather Weather Weather R33 4:35 4:55 4:35 4:40 4:35 4:35 4:50 5:10 4:35 4:45 5:10 4:35 4:35 5:25

 ,C,D,L,M,R R34 4:35    4:35  4:35    4:35   4:35   4:35      4:35 4:35  4:35      4:35 4:35  4:35      4:35       4:35 CLM,O R35 4:35    4:35  4:35    4:35   4:35   4:35      4:35 4:35  4:35      4:35 4:35  4:35      4:35       4:35
 ,C,L,M,N,0)

R36 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 (B,C,L,M,N R37 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:30 A,B,C,L,M,N R38 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 4:35 5:30 AB,C,L,M R39 4:40 4:55 4:35 4:45 4:35 4:40 5:05 5:15 4:35 4:45 5:10 4:35 4:35 5:55 4:35 4:55 4:35 4:45 4:35 4:50 4:50 5:10 4:35 4:45 5:10 4:35 4:35 5:25 (A,B,C,D,L,M, 4:55 5:00 4:35 4:45 4:35 5:00 5:10 5:20 4:35 4:50 5:10 4:35 4:35 5:55 N,O,R J-20 Rev. 18-1 January, 2018

Figure J-5 Evacuation Route map for MNS

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J-21 Rev. 18-1 January, 2018

Figure J-6 Summary of Population and Demand Transit- Special Shadow External Sub-Zone Residents Transients Employees Schools Total Dependent Facilities Population Traffic A 18433 335 1478 3632 0 338 24,216 B 931 17 0 2231 0 896 4,075 C 1484 27 0 0 0 0 1,511 D 22994 418 2065 1163 102 4900 31,642 E 37228 674 472 2354 0 4134 44,862 F 30364 552 7850 14154 740 6591 60,251 G 25408 462 1150 3311 149 5356 35,836 H 9665 176 694 2604 0 4826 17,965 I 8053 146 19 121 0 0 8,339 J 7447 135 128 1398 107 1471 10,686 K 2272 41 111 0 0 0 2,424 L 1247 23 13 0 0 0 1,283 M 238 4 0 0 0 0 242 N 5381 98 134 1079 0 509 7,201 0 3705 67 0 501 0 0 4,273 p 10377 189 5232 1048 112 3408 20,366 Q 3394 62 13 56 0 0 3,525 R 1667 30 0 43 0 0 1,740 s 14970 272 0 756 236 1652 17,886 Shadow 9085 61906 70,991 Region Total 205,258 3,728 19,359 34,451 1,446 43,166 61,906 369,314 NOTE: Shadow population has been reduced to 20%. NOTE: Shadow Facilities include both medical facilities and correctional facilities. J-22 Rev. 18-1 January, 2018

K. RADIOLOGICAL EXPOSURE CONTROL To assure that means for controlling radiological exposures in an emergency are established for emergency workers. K.1 Onsite Exposure Guidelines Onsite exposure guidelines consistent with EPA 400-R-92-001, Table 2-2, "Guidance on Dose Limits for Workers Performing Emergency Services" are shown in Figure K-1. Members of outside emergency services responding to a call from the site are considered emergency workers and must also be protected from excessive radiation doses. Their doses are not to exceed guidelines as established in Figure K-1. K.2 Doses in Excess of 10CFR Part 20 The Emergency Coordinator is responsible for authorizing emergency workers to receive doses in excess of 10CFR20 limits. An on-site radiation protection program shall be implemented during emergencies which shall be consistent with ALARA conditions. The site will be responsible for providing medical treatment and rescue efforts for life-saving missions. Site procedures are in place for expeditious decision-making with reasonable consideration of the relative risks involved in a lifesaving mission involving radiation exposure. K.3 Emergency Personnel Exposure and Records K.3.a Distribution of Dosimetry Provisions have been made for maintaining records of emergency personnel during a radiological emergency on a 24-hour per day basis. The Operations Support Center will provide a means for keeping track of exposure to personnel involved in a radiological accident. Distribution of dosimeters (self-reading and TLD badges) will be provided for all personnel. The issuance of High Range and/or Multiple Dosimetry will be in accordance with Radiation Protection procedures. Should any offsite agency respond to an emergency at the site during a nuclear emergency, dosimeters will be provided for their use to determine any exposure. Rev. 18-1 K-1 January, 2018

K.3.b Dose Records The Operations Support Center through the Radiation Protection section shall have the responsibility of keeping records of the doses received by emergency personnel involved in any radiological accident. Normal operating procedures shall be followed for the use of dosimeters and the TLD badges. Distribution of the dosimeters and badges shall be through Radiation Protection. K.4 State/Local Plan for Authorizing Doses Exceeding PAG's See County and State Plans. K.5 Decontamination K.5.a Action Levels For Determining the Need For Decontamination Guidelines as established in the System Radiation Protection Procedures will be used to determine action levels for decontamination. Pre-planning efforts have been established by the Radiation Protection Section. K.5.b Radiological Decontamination AD-RP-ALL-2009, Personnel Contamination Monitoring and Reporting, defines the specific action levels for determining the need for decontamination of personnel. PTIOIAl4600/088, Functional Check of Emergency Vehicle and Equipment, defines the means for availability of supplies, instruments and equipment. Radiation Protection procedures provides direction for waste disposal. HP/O/B/1009/024, Personnel Monitoring for Emergency Conditions *promotes means for decontamination of emergency personnel. Handling of contaminated injured personnel is provided in HP/O/B/1009/022, Accident and Emergency Response. Transportation of contaminated injured personnel is described in site procedure RP/O/A/5700/005, Care and Transportation of Injured Individual(s) From Site to Offsite Medical Facility. K.6 Contamination Control Measures K.6.a Area Access Control - The site will be evacuated when site management declares a Site Evacuation and a potential threat exists for safety of non-essential personnel. Once the site has been evacuated, access to the site will be limited by the Highway Patrol on the public highway and then Site Security will limit access to the site except through established access procedures. K.6.b Drinking Water and Food Supplies - Drinking water and food supplies can be brought in by private vendor if necessary. Arrangements will be made by the Commodities and Facilities Manager/Designee. Rev. 18-1 K-2 January, 2018

K.6.c Recovery efforts will be determined by the Emergency Operations_ Facility Organization (see Section M). K.7 Decontamination of Personnel at Relocation Assembly Area Should non-essential plant personnel be evacuated from site to a relocation area, provisions for extra clothing and decontaminants suitable for any type of contamination have been made. Extra clothing and supplies have been placed at the relocation site to take care of site personnel. Relocation assembly areas have been determined so that site personnel can be relocated to a safe site quickly and can be decontaminated (if necessary), monitored and released. Records will be made of the exposure of all personnel released from the relocation site. (Site procedures provide for emergency supplies to be provided at the off-site relocation assembly area.) Rev. 18-1 K-3 January, 2018

FIGUREK-1 Emergency Worker Exposure Guidelines (a) Dose Limits Activity Total Effective Dose Lens of Eye Other Organs (b) Eguivalent (TEDE) All 5rem 15 rem 50rem Protecting Valuable lOrem 30rem 100 rem Property Lifesaving or 25rem 75rem 250rem Protection of Large Populations Lifesaving or >25rem > 75 rem >250rem Protection of Large

    .Populations (c)

(a) Excludes declared pregnant women (b) Includes skin and body extremities (c) Only on a volunteer basis to persons fully aware of the risks involved Based on EP A-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents Rev. 18-1 K-4 January, 2018

L. MEDICAL AND PUBLIC HEALTH SUPPORT L.l Hospital and Medical Support Hospitals - Carolinas Medical Center; Charlotte, NC - (Agreement #1 App. 5) Medical Support - Local - Carolinas Medical Center (Agreement #1 App. 5) Backup - Oak Ridge National Lab Hospital, Oak Ridge, Tenn. (Agreement #13, App. 5) Ambulance Service Mecklenburg Emergency Medical Services Agency (MEDIC), Charlotte, NC (Agreement #6 App. 5) L.2 On-site First-Aid Capability The on-site medical facilities include two First Aid areas and a bioassay facility. One First Aid facility, located outside of the protected area southeast of the Administration Building, is used for the treatment of persons injured, outside the Protected Area, in accidents or emergencies not involving radioactive contamination. This facility is equipped with a sink, a treatment chair, lavatory, a wheeled stretcher, a resuscitator, and medical/first aid supplies. The second First Aid area, located in the Auxiliary Building near the Radiation Protection office area, is used for treatment of persons injured in accidents or emergencies involving radiological contamination. This facility has a decontamination area with a shower, a treatment table and medical/first aid supplies. The bioassay facility, located in the Administration Building, is used in emergencies to determine if personnel have inhaled or ingested radioactive materials, or if such materials have entered wounds or been absorbed through the skin. The bioassay facility is equipped with a shielded body-burden analyzer and a thyroid-burden analyzer; liquid scintillation counting capabilities for tritium analyses are available in the Radiation Protection area and laboratory in the Radiation Control Area. L.3 Public, Private, Military Hospitals; Emergency Medical Facilities See State of North Carolina FNF Plans. Rev. 18-1 L-1 January, 2018

L.4 Transport of Accident Victims McGuire Nuclear Site has agreements with the Carolinas Medical Center, AND MEDIC to provide transportation for any medical emergency patient (may or may not be contaminated). If contaminated, efforts will be made to decontaminate the victim before transportation as long as the decontamination does not obstruct the medical attention given the victim or cause an unnecessary delay in transporting. During transportation Radiation Protection personnel will accompany the victim and prevent the further spread of contamination using procedure RP/O/A/5700/005, "Care and Transportation of Contaminated Injured Individuals". Any item(s) found to be contaminated after the treatment of a contaminated patient at the Carolinas Medical Center or any other medical facility will be decontaminated or replaced by Duke Energy. Rev. 18-1 L-2 January, 2018

M. RECOVERY AND REENTRY PLANNING AND POST ACCIDENT OPERATIONS M.1 Recovery/Reentry Plans and Procedures In any site emergency involving radioactive contamination or other emergency condition, the immediate action is directed to limiting the consequences of the incident in a manner that will afford maximum protection to the public. Once the immediate protective actions have established an effective control over the incident, the emergency actions will shift into the recovery phase. AD-EP-ALL-0110, Recovery, provides guidance in establishing the recovery/reentry organization and actions. The EOF Director will inform members of the response organization that a recovery operation is to be initiated and inform them of any changes in the organization that may occur. Implementation of Recovery Operations would occur as follows:

1. Termination of General Emergency or Site Area Emergency

2. Deescalation to Non-Emergency Condition

3. Activation of Recovery Organization The emergency is not considered to be over until Duke Energy, NRC and the states agree that the public is afforded comparable safety assurance to that which exists during periods of normal station operation. Specifically:

1. Radiation levels in site areas are stable or decreasing with time.

2. Releases of radioactive materials to the environment from the site are under control or have ceased.

3. Any fire, flooding or similar emergency conditions are controlled or have ceased.

Public officials are kept informed of recovery plans so that they can properly carry out their responsibilities to the public. Periodic briefings of media representatives are held to inform the public of recovery plans and progress made. Periodic status reports are given to company employees at other locations and to government and industry representatives. M.l.a Outline of Site Recovery Plans A Recovery Plan will be developed, based on event, to provide a plan for recovery from and return to an operational status following a notification of Unusual Event, Alert, Site Area Emergency or a General Emergency. Rev. 18-1 M-1 January, 2018

The plans and procedures for site reentry will consider existing as well as potential conditions inside containment. Prior to reentry, the following shall be addressed:

1. Review all available radiation survey data.

2. Determine site areas potentially affected by radiological hazards.

3. Review radiation exposure history of all personnel scheduled to participate in recovery operations. Determine the need for additional personnel.

4. Review the adequacy of radiation survey equipment available. Determine the need for additional equipment and a source of procurement.

5. Pre-plan team activities, including areas to be surveyed, anticipated radiation levels, survey equipment required, protective clothing requirements, access control procedures, exposure control procedures and communication capabilities.

6. Conduct comprehensive radiation survey of site facilities and define all radiological problem areas.

7. Isolate and post with appropriate warning signs all "high radiation areas" and areas of contamination.

8. Perform visual inspection of site areas and equipment.

9. All radiological conditions discovered and existing in the facility as determined by the re-entry survey will be evaluated by site management.

10. Upon evaluation of the radiological condition, site management will determine what procedures are required to restore the site to a normal status.

11. Personnel radiation exposure will be closely controlled and documented.

12. Recovery coordinators will take appropriate actions to ensure that emergency personnel and equipment leaving the Radiation Control Area are not contaminated, that radiological conditions at the scene of the emergency are properly defined, barricaded and posted with appropriate signs.

M.l.b Outline of Recovery Plans Recovery from an emergency is guided by the following principles: Rev. 18-1 M-2 January, 2018

a. The protection of the public health and safety is* the foremost consideration in formulating recovery plans.

b. Public officials would be kept informed of recovery plans so that they can properly carry out their responsibilities to the public.

c. Periodic information would be provided to the news media so that they can provide information to the public regarding recovery plans and progress made.

d. Periodic status reports would be given to company employees at other locations and to government and industry representatives.

e. The radiation doses to employees and other radiation workers would be kept as low as reasonably achievable.

M.2 Recovery Organization Before entering the Recovery phase, the EOF Director/Emergency Coordinator should establish a Recovery organization that is appropriate for the existing on-site and off-site conditions. Figure M-1 describes a suggest~d organization structure. It may be modified or supplemented as necessary to fit the particular circumstances. In some situations (such as no core damage), the normal organization may be adequate and a Recovery Organization may not be needed. The recovery activities would be managed much like a normal outage, except that certain activities unique to the post accident situation may be managed by the Recovery organization. The organization would :function as a matrix management organization to coordinate activities with the nonnal company organization. This organization may be located at the Emergency Operations Facility or the site, as appropriate. The primary positions in the Recovery Organization are described below: Recovery Manager - Overall management of recovery activities. Coordination with Federal, State, and local government agencies Onsite Recovery Director - Directs the recovery activities onsite to restore the plant to pre-incident conditions. Offsite Recovery Director - Directs interface with Federal, State and local agencies during the recovery process. Radiological Assessment Manager (if needed) - Coordinates radiological and environmental assessment with Federal and State agencies. Coordinates offsite radwaste management and decontamination activities. Rev. 18-1 M-3 January, 2018

Company Spokesperson - Manages communications of recovery activities. Informs the news media, employees, etc. Other Support - Individuals from other company or outside groups may be assigned to specific recovery positions or to perform specific tasks based on recovery needs. EOF Services Manager or Nuclear supply Chain Manager - Coordinates activities such as purchasing, finance, insurance, human resources, transportation, etc. Other site management and supervisory personnel will interface with recovery operations as I necessary and as warranted. M.3 Information to Members of Recovery Organization The EOF Director will take the following steps to inform members of the Emergency Operations Facility, Site Organization, and off-site agencies that Recovery Operations are being initiated and that activities associated with bringing the plant to a safe shutdown condition are completed:

1. Develop a brief message as to the time and date of Recovery Operations initiation as well as any necessary organizational realignments.

2. Distribute the message to EOF Managers, News Manager, Emergency Coordinator, state and local officials, NRC and other representatives. Ask that each person inform those under his/her direction.

M.4 Total Population Exposure Estimates The Radiological Assessment Manager (or successor in Recovery/ Reentry Operations) will periodically update the estimate of total population exposure. See Section 1.10. Rev. 18-1 M-4 January, 2018

FIGUREM-1 RECOVERY ORGANIZATION Recovery Manager I I I I I Onsite Offsite Company Other Recovery Recovery Spokesperson Support Director Director Directs Recovery Directs the interface Directs the Public Other individuals or activities to restore the between Duke Energy Information Program groups assigned specific plant to pre-incident and Federal, State and during the Recovery tasks to support activities conditions or permanent Local Agencies during Phase. during the Recovery safe shutdown the Recovery Phase. Phase, such as Legal, Financial, Engineering, etc. Rev. 18-1 M-5 January, 2018

P. RESPONSIBILITY FOR THE PLANNING EFFORT To assure that responsibilities for plan development, review and distribution of emergency plans are established and that the Emergency Planning staff are properly trained. P.1 Emergency Planning Staff Training Emergency Planning Group personnel will attend training/workshops, information exchange meetings with other licensees, and conferences held by industry and government agencies, as available, to maintain current knowledge of the overall planning effort. The Emergency Preparedness Manager is required to attend offsite training on an annual basis. This training will be documented in site Emergency Planning files or the Training group database (i.e. People Soft, etc.). P.2 Emergency Response Planning The Site. Vice P:esident h_as the overall authority and responsibility for the Site Emergency I Plan. This planning effort 1s delegated to the Emergency Preparedness Manager.

P.3 Site Emergency Preparedness Manager The Emergency Preparedness Manager has the overall authority and responsibility for site emergency response planning as well as the responsibility for the development and updating of the site Emergency Plan and coordination of this plan with other response organizations.

P.4 Review ofEmergencyPlan Review and updating of the site Emergency Plan and Emergency Plan Implementing Procedures shall be certified to be current on an annual basis. Any changes identified by drills and exercises shall be incorporated into the Site Emergency Plan. On an annual basis, the Emergency Preparedness Manager will provide each state and local organization responsible for off-site activation and protective action decision-making, a copy of the nuclear site procedures appropriate for their area on emergency classification and notification. A response will be requested by letter within 30 days that a review has been completed with concurrence with the EAL's used for event classification and for protective action recommendations. If problem areas are identified, the Emergency Preparedness Manager will ensure resolution. P.5 Distribution of Revised Plans The Emergency Plan and approved changes shall be forwarded to individuals and organizations listed in App. 6, Revised pages shall be dated and marked to show where changes have been made. Rev. 18-1 P-1 January, 2018

P.6 Supporting Plans Figure P-1 gives a detailed listing of supporting plans to the McGuire Nuclear Site Emergency Plan. P.7 Implementing Procedures Written procedures will be established, implemented, and maintained covering the activities associated with emergency plan implementation. Each procedure, and changes thereto, shall be reviewed and approved by the responsible implementing manager prior to implementation and shall be reviewed periodically as set forth in administrative procedures. McGuire Emergency Plan Implementing Procedures are listed in Figure P-2 with a reference to the section of Emergency Plan implemented by each procedure. Figure P-3 contains the distribution list for McGuire Emergency Plan Implementing Procedures. P.8 Table of Contents The McGuire Nuclear Site Emergency Plan contains a specific table of contents. The McGuire Nuclear Site Emergency Plan has been cross referenced to the applicable sections ofNUREG-0654 Rev. 1. P.9 Audit of Emergency Plan The Nuclear Oversight (NOS) Audit Manager will arrange for an independent review of McGuire Nuclear Station's Emergency Preparedness Program at a frequency specified in 10 CPR 50.54(t). NOS will audit the Plan, Emergency Plan Implementing Procedures, Training, Drills and Exercise, facilities and equipment for conformance with 10 CPR 50.47, 10 CPR 50.54, and 10 CFR 50 Appendix E. The independent review will include the following plans, procedures, training programs, drills/exercises, equipment, and State/local government interfaces:

1. McGuire Nuclear Site Emergency Plan and Implementing Procedures

2. State/Local Support Agency Training Program

3. Emergency Response Training Program

4. Public and Media Training/Awareness

5. Equipment - Communications, Monitoring, Meteorological, Public Alerting

6. State/Local Plan Interface Rev. 18-1 P-2 January, 2018

The review findings will be submitted to the appropriate corporate and nuclear site management. Appropriate portions of the review findings will be reported to the involved federal, state, and local organizations. The corporate or nuclear site management, as appropriate, will evaluate the findings affecting their area of responsibility and ensure effective corrective actions are taken. The result of the review, along with recommendations for improvements, will be documented and retained for a period of five years. P .10 Telephone Number Updates Telephone numbers reflected in the online organization charts will be updated quarterly in accordance with PT/O/A/4600/091, Periodic Test of Technical Support Center Communications and Supplies. Rev. 18-1 P-3 January, 2018

DUKE ENERGY MCGUIRE NUCLEAR SITE FIGUREP-1 SUPPORTING PLANS

1. North Carolina Emergency Response Plan in support of McGuire Nuclear Site

2. South Carolina Operational Radiological Emergency Response Plan in support of Fixed Nuclear Facilities (McGuire Nuclear Site)

3. Iredell County, N.C., Radiological Emergency Response Plan in Support of the McGuire Nuclear Site

4. Mecklenburg County, N.C., Radiological Emergency Response Plan in Support of the McGuire Nuclear Site

5. Gaston County, N.C., Radiological Emergency Response Plan in Support of the McGuire Nuclear Site

6. Lincoln County, N.C., Radiological Emergency Response Plan in Support of the McGuire Nuclear Site

7. Catawba County, N.C., Radiological Emergency Response Plan in Support of the McGuire Nuclear Site

8. Cabarrus County, N.C., Radiological Emergency Response Plan in Support of the McGuire Nuclear Site

9. Emergency Response Plan, Water Reactors Division, Westinghouse Electric Corporation

10. N.R.C. Region II Incident Response Plan

11. Interagency Radiological Assistance Plan - Region 3 - U.S. Department of Energy

12. INPO Emergency Response Plan Rev. 18-1 P-4 January, 2018

MCGUIRE FIGUREP-2 PAGE 1 OF4 EMERGENCY PLAN IMPLEMENTING PROCEDURES Procedure# Title Emergency Plan Section Irn2lemented AP/O/A/5500/047 Security Events (Proprietary Info) Section J RP/0/A/5700/000 Classification of Emergency Section D, E, I RP/O/A/5700/001 Notification of Unusual Event Section D, E, 1.1, J.7 RP/0/A/5700/002 Alert Section D, E, 1.1, J.7 RPIOIAl5700/003 Site Area Emergency Section D, E, 1.1, J.7, M.1 RP/O/A/5700/004 General Emergency Section D, E, 1.1, J.7, M.1 RP/0/A/5700/006 Natural Disasters SectionD RP/0/A/5700/007 Earthquake Section D, H.6 RP/0/A/5700/008 Release of Toxic or Flammable Gases Section D RP/0/A/5700/09 Collisions/Explosions Section D RP/O/A/5700/010 NRC Immediate Notification SectionD RP/O/A/5700/011 Conducting a Site Assembly, Site Evacuation or Section E.2, J, K.7 Containment Evacuation Rev. 18-1 P-5 January, 2018

MCGUIRE FIGUREP-2 PAGE2 OF4

                      - EMERGENCY PLAN IMPLEMENTING PROCEDURES Procedure#                             Title                              Emergency Plan Section Imglemented RP/O/A/5700/019           Core Damage Assessment                             Section I RP/O/A/5700/022           Spill/Incident Response Procedure                  Appendices 7, 8, 9 RP/O/A/5700/026           Operations/Engineering Required Actions in the Technical Support Center (TSC)

RP/O/B/5700/029 Notifications to Offsite Agencies from the Control Section E Room HP /0/B/1009/002 Alternative Methods for Determining Dose Rate Section D, I.6 within the Reactor Building HP /0/B/1009/003 Recovery Plan Section M HP /0/B/1009/006 Procedure for Quantifying High Level Gaseous Section D, I.3 Radioactivity Release During Accident Conditions HP /0/B/1009/010 Releases of Liquid Radioactive Materials Section D, I.3 Exceeding Selected Licensee Commitments HP/0/B/1009/021 Estimating Food Chain Doses Under Post-Accident I.10 Conditions HP /0/B/1009/022 Accident and Emergency Response Section I, Section E HP /0/B 1009/024 Personnel Monitoring for Emergency Conditions J.3, K.7 Rev. 18-1 P-6 January, 2018

MCGUIRE FIGUREP-2 PAGE3 OF4 EMERGENCY PLAN IMPLEMENTING PROCEDURES Procedure# Title Emergency Plan Section Imglemented AD-EP-ALL-0101 Emergency Classification Section D, E, I AD-EP-ALL-0103 Activation and Operation of the EOF Section B, H, M.1 AD-EP-ALL-0104 ERO Common Guidelines and Forms Section B, E, F, G, I, J, K, M AD-EP-ALL-0105 Activation and Operation of the TSC Section B, H, M.1 AD-EP-MNS-0105 MNS Site Specific TSC Support SectionH AD-EP-ALL-0106 Activation and Operation of the OSC SectionH AD-EP-MNS-0106 MNS Site Specific OSC Support SectionH AD-EP-ALL-0108 Joint Information System Support Section G AD-EP-ALL-0109 Protective Action Recommendations J.7 AD-EP-ALL-0110 Recovery SectionM AD-EP-ALL-0202 Emergency Response Offsite Dose Assessment Section I AD-EP-ALL-0203 Field Monitoring During Declared Emergency Section E, I.7, I.8, I.9 AD-EP-MNS-0203 MNS Site Specific Field Monitoring Section E, I.7, I.8, I.9 AD-EP-ALL-0204 Distribution of Potassium Iodide Tablets in the J.6 Event of a Radioiodine Release AD-EP-ALL-0205 Emergency Exposure Controls Section K.2 AD-EP-ALL-0304 State and County Notifications Section E Rev. 18-1 P-7 January, 2018

MCGUIRE FIGUREP-2 PAGE40F4 EMERGENCY PLAN IMPLEMENTING PROCEDURES Procedure# Title Emergency Plan Section Im12lemented EP Manual Section 1.1 Emergency Organization Sections B, E, H PTIOIA/46001088 Functional Check of Emergency Vehicle and Section H.11 Equipment Rev. 18-1 P-8 January, 2018

FIGUREP-3 McGUIRE NUCLEAR SITE EMERGENCY PLAN IMPLEMENTING PROCEDURES DISTRIBUTION Control No.

2. Radiation Protection Manager

3. Emergency Preparedness Manager, Oconee

4. McGuire Nuclear Training

5. Operations StaffManager

6. Site Emergency Planner, MGOlEP

7. NRC Site Representative, McGuire Nuclear Site (forwarded by McGuire Emergency Planning)

8. Operator Training Director

13. Emergency Preparedness Manager, CNS

14. Director, Division of Radiation Protection

16. NCEM REP Program Manager

17. Tina Kuhr, Fleet Emergency Planning

19. Emergency Operations Facility, EOF Director's Area (MNS Emergency Planning, custodians)

20. McGuire Nuclear Site, Document Control

21. NCEM Western Branch Office Manager

22. NRC Document Control Desk, Washington D.C. (forwarded 1 copy by McGuire Emergency Planning)

Rev. 18-1 P-9 January, 2018

FIGUREP-3 McGUIRE NUCLEAR SITE EMERGENCY PLAN IMPLEMENTING PROCEDURES DISTRIBUTION Control No.

23. NRC, Regional Administrator, Atlanta, GA (forwarded 1 copy by McGuire Emergency Planning)

24. NRC, Regional Administrator, Atlanta, GA (forwarded 1 copy by McGuire Emergency Planning)

25. NRC Office of Nuclear Materials Safety and Safeguards Rev. 18-1 P-10 January, 2018

APPENDIX 1 1.0 DEFINITIONS AFFECTED PERSONS Persons who have received radiation exposure or have been physically injured as a result of an accident to a degree requiring special attention as individuals, e.g., decontamination, first aid or medical services. ANNUAL For periodic emergency planning requirements, annual is defined as twelve months, with a maximum interval of 456 days. ASSESSMENT ACTION Those actions taken during or after an accident to obtain and process information that is necessary to make decisions to implement specific emergency measures. BIENNIAL For periodic emergency planning requirements, biennial is defined as at least once every two years, with a maximum interval of912 days. (Note that this does not apply to the scheduling of biennial exercises. An exercise can occur at any time during the second calendar year after the previous exercise.) CORRECTIVE ACTIONS Emergency measures taken to ameliorate or terminate an emergency situation at or near the source of the problem to prevent an uncontrolled release of radioactive material or to reduce the magnitude of the release, e.g., shutting down equipment, fire-fighting, repair and damage control. DRILL A drill is a supervised instruction period aimed at testing, developing, and maintaining skills in a particular operation. EMERGENCY ACTION LEVELS (EAL's) A pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given emergency class. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, or another phenomenon which, if it occurs, indicates entry into a particular emergency class. EMERGENCY OPERATIONS FACILITY (EOF) The Emergency Operations Facility is the facility utilized for direction and control of all emergency and recovery activities with emphasis on the coordination of off-site activities such as dispatching mobile emergency monitoring teams, communications with local, state and federal agencies, and coordination of corporate and other outside support. 1 Rev. 18-1 January, 2018

EMERGENCY PLANNING ZONE (EPZ) The area for which planning is needed to assure that prompt and effective actions can be taken to protect the public in the event of an accident. The plume exposure EPZ is about 10 miles in radius and the ingestion exposure EPZ is about 50 miles in radius. EXCLUSION AREA The nuclear site property out to a radius of 2500 feet, that meets the 10CFRl 00 definition. EXERCISE An exercise is an event that tests the integrated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations. INGESTION EXPOSURE PATHWAY The principle exposure from this pathway would be from ingestion of contaminated water or foods such as milk or fresh vegetables. The time of potential exposure could range in length from hours to months. MONTHLY For periodic emergency planning requirements, monthly is defined as once each month, with a maximum interval of 3 8 days. OPERATIONAL SUPPORT CENTER (OSC) In the event of an emergency, shift support personnel (e.g., auxiliary operators and technicians) other than those required and allowed in the control room shall report to this center for further orders and assignment. PLUME EXPOSURE PATHWAY The principle exposure sources from this pathway are (a) external exposure to gamma radiation from the plume and from deposited material and (b) inhalation exposure from the passing radioactive plume. The time of potential exposure could range from hours to days. POPULATION-AT-RISK Those persons for whom protective actions are being or would be taken. PROTECTED AREA An area encompassed by physical barriers and to which access is controlled. PROTECTNE ACTIONS Those emergency measures taken after an uncontrolled release of radioactive materials has occurred, for the purpose of preventing or minimizing radiological exposures to persons that would be likely to occur if the actions were not taken. 2 Rev. 18-1 January, 2018

PROTECTIVE ACTION GUIDES (PAG) Projected radiological dose or dose-commitment values to individuals in the general population that warrant protective action following a release of radioactive material. Protective actions would be warranted provided the reduction in individual dose expected to be achieved by carrying out the preventive action is not offset by excessive risks to individual safety in taking the protective action. The PAG does not include the dose that has unavoidably occurred prior to the assessment. QUARTERLY For periodic emergency planning requirements, quarterly is defined as once every three months, with a maximum interval of 112 days. RECOVERY ACTIONS Those actions taken after the emergency to restore affected property as nearly as practicable to its pre-emergency condition. SEMI-ANNUAL For periodic emergency planning requirements, semi-annual is defined as once every 6 months, with a maximum interval of 228 days.

SITE That part of the nuclear site property consisting of the Reactor, Auxiliary, Turbine, Service Buildings and grounds, contained within the outer security area fence.

TECHNICAL SUPPORT CENTER (TSC) This on-site center is for use by plant management, technical and engineering support personnel. In an emergency, this center shall be used for assessment of plant status and potential off-site impact in support of the control room command and control function. TRIENNIAL For periodic emergency planning requirements, triennial is defined as at least once every three years, with maximum interval of 1369 days. VITAL AREA Areas within the Protected Area that house equipment important for nuclear safety. Access to a Vital Area is allowed only if an individual has been authorized to be in that area per the Security plan, therefore Vital Area is a Security term. WEEKLY For periodic emergency planning requirements, weekly is defined as once every 7 days, with a maximum interval of 9 days. 3 Rev. 18-1 January, 2018}}

MNS-18-009, Site Emergency Plan

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4.0 REFERENCES

4.0 REFERENCES

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