Text

Emergency Plan, Revision 76, On-Shift Staffing Analysis, Revision 2, and Emergency Plan Implementing Procedures
	ML21033A513
Person / Time
Site:	Cooper
Issue date:	01/06/2021
From:	; Nebraska Public Power District (NPPD)
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML21033A511	List: ML21033A512; NLS2021001, Emergency Plan, Revision 76, On-Shift Staffing Analysis, Revision 2, and Emergency Plan Implementing Procedures;
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	NLS2021001
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{{#Wiki_filter:NLS2021001 Attachment Page 1 of 2 ATTACHMENT Cooper Nuclear Station Report of Change and Summary of 5054(q) Analyses Emergency Plan, Revision 76, Emergency Plan Implementing Procedure 5.7.1, Revision 67, Emergency Plan Implementing Procedure 5.7.1.1, Revision 0, On-Shift Staffing Analysis, Revision 2 EPIP 5.7.1, Revision 67 EPIP 5.7.1.1, Revision 0 Change

Description:

Guidance for initial classification, declaration, reclassification, and misclassification was deleted from Emergency Plan Implementing Procedure (EPIP) 5. 7.1, Emergency Classification, and was relocated to new EPIP 5. 7 .1.1, Emergency Classification Process. Emergency Directors will use this new EPIP to assess plant conditions, maintain oversight of their emergency response facility, obtain independent reviews of emergency action levels, and conduct joint discussions with other emergency response facilities for classification prior to declaring an emergency event. EPIP 5. 7 .1 continues to provide the formal set of threshold conditions necessary to classify an event as required per 10 CFR 50.47(b)(4). EPIP 5.7.1.1 is not considered part of the Cooper Nuclear Station (CNS) Emergency Plan as it does not include content meeting the requirements for inclusion in the Emergency Plan per 10 CFR 50.4 7 or Regulatory Guide 1.219. Change Summary of Analysis (10 CFR 50.54(q) Screen): The emergency planning impact screen determined the changes did not impact a planning standard or program element and as such, did not require completion of a full 10 CFR 50.54(q) evaluation. Emergency Plan, Revision 76 Change

Description:

Emergency Plan, Appendix A, was revised to add new EPIP 5.7.1.1 to the list of implementing procedures as well as a brief description of the procedure.

NLS2021001 Attachment Page 2 of2 Change Summary of Analysis (10 CFR 50.54(q) Screen): The emergency planning impact screen determined the changes did not impact a planning standard or program element and as such, did not require completion of a full 10 CFR 50.54(q) evaluation. On-Shift Staffing Analysis, Revision 2 Change

Description:

Section I of the CNS On-Shift Staffing Analysis was revised to add a note requiring revisions to the document to be processed per the station procedure for license basis document changes since the On-Shift Staffing Analysis is considered part of the CNS Emergency Plan. Change Summary of Analysis (10 CFR 50.54(q) Screen): The emergency planning impact screen determined the change did not impact a planning standard or program element and as such, did not require completion of a full 10 CFR 50.54(q) evaluation.

NLS2021001 Page 1 of 193 ENCLOSURE 1 Cooper Nuclear Station Emergency Plan, Revision 76

I I { NEBRASKA PUBLIC POWER DISTRICT EMERGEN CY PLAN FOR COOPER NUCLEAR STATION EP staff KiJ~ REVISED BY REVIEWED BY APPROVED BY ~ D a t e 12/15/2020 I EMERGENCY PLAN REVISION 76 PAGE 1 OF 192 I

1. DEFINITIONS .....................................................................................................................8

2. SCOPE AND APPLICABILITY ......................................................................................... 10

3.

SUMMARY

OF THE NEBRASKA PUBLIC POWER DISTRICT (NPPD) CNS EMERGENCY PLAN ........................................................................................................11

4. EMERGENCY CONDITIONS ........................................................................................... 13 4.1 CLASSIFICATIONS .............................................................................................. 13 4.1.1 NOTIFICATION OF UNUSUAL EVENT .......................................... 14 4.1.2 ALERT ............................................................................................. 14 4.1.3 SITE AREA EMERGENCY .............................................................. 15 4.1.4 GENERAL EMERGENCY ............................................................... 15 4.2 OFF-SITE RADIOLOGICAL ASSESSMENT ........................................................ 16 4.3 SPECTRUM OF POSSIBLE ACCIDENTS AND INITIATING EVENTS ................ 17

5. ORGANIZATIONAL CONTROL OF EMERGENCIES ...................................................... 32 5.1 NORMAL OPERATING ORGANIZATION ............................................................ 32 5.1.1 LINES OF AUTHORITY ................................................................... 32 5.1.2 RESPONSIBILITIES/FUNCTIONS .................................................. 33 5.2 EMERGENCY RESPONSE ORGANIZATION ...................................................... 34 5.2.1 EMERGENCY DIRECTOR .............................................................. 35 5.2.2 TSC DIRECTOR .............................................................................. 36 5.2.3 EOF DIRECTOR ..............................................................................38 5.2.4 OPERATIONAL SUPPORT CENTER (OSC) SUPERVISOR. ........ .40 5.3 OFF-SITE EMERGENCY ORGANIZATION ............................ :........................... .41 5.3.1 - JOINT INFORMATION CENTER (JIC) ........................................... .41 5.3.2 PUBLIC INFORMATION SUPPORT .............................................. .42 0 5.3.3 5.4 CONTRACT SUPPORT ..................................................................43 PARTICIPATING FEDERAL, STATE AND LOCALAGENCIES ........................... 44 5.4.1 THE STATE OF NEBRASKA .......................................................... 44 5.4.2 THE STATE OF MISSOURI ........................................................... .45 5.4.3 THE STATES OF KANSAS/IOWA. ................................................. .45 5.4.4 NUCLEAR REGULATORY COMMISSION .................................... .46

6. EMERGENCY MEASURES .............................................................................................54 6.1 SITE EMERGENCY ALARMS ..............................................................................54 6.2 NOTIFICATION AND ACTIVATION OF EMERGENCY RESPONSE ORGANIZATIONS ................................................................................................54 6.2.1 ON-SITE PLANT PERSONNEL. ...................................................... 54 6.2.2 OFF-SITE PLANT PERSONNEL ..................................................... 54 6.2.3 JOINT INFORMATION CENTER (JIC) ............................................ 55 6.2.4 OFF-SITE AUTHORITIES AND SUPPORT AGENCIES ................. 55 6.2.5 NUCLEAR REGULATORY COMMISSION (NRC) .......................... 55 6.3 ASSESSMENT ACTIONS .....................................................................................56 6.3.1 POST-ACCIDENT SAMPLING SYSTEM ........................................ 56 6.3.2 METEOROLOGICAL DATA ............................................................ 56 6.3.3 DOSE ASSESSMENT ..................................................................... 56 6.4 CORRECTIVE ACTIONS ......................................................................................57 6.5 PROTECTIVE ACTIONS ......................................................................................58 6.5.1 RESCUE OPERATIONS ................................................................. 60 6.5.2 ON-SITE PROTECTIVE EQUIPMENT AND SUPPLIES ................. 60 6.5.3 PERSONNEL ASSEMBLY AND ACCOUNTABILITY ...................... 61 6.5.4 DISMISSAL AND EVACUATION ..................................................... 61 EMERGENCY PLAN REVISION 76 PAGE 2 OF 192

6.5.5 CONTAMINATION AND DOSE CONTROL MEASURES ............... 61 6.5.6 SECURITY AND ACCESS CONTROL. ........................................... 62 6.6 AID TO AFFECTED PERSONNEL ....................................................................... 62 6.6.1 EMERGENCY PERSONNEL DOSE CRITERIA .............................. 62 6.6.2 DECONTAMINATION AND FIRST AID ........................................... 63 6.6.3 MEDICAL TRANSPORTATION ....................................................... 63 6.6.4 MEDICAL TREATMENT FACILITIES .............................................. 64

7. EMERGENCY RESPONSE FACILITIES AND EQUIPMENT ........................................... 68 7.1 CONTROL ROOM ................................................................................................68 7.2 EMERGENCY RESPONSE FACILITIES .............................................................. 68 7.2.1 TECHNICAL SUPPORT CENTER .................................................. 68 7.2.2 OPERATIONAL SUPPORT CENTER ............................................. 69 7.2.3 EMERGENCY OPERATIONS FACILITY ........................................ 69 7.2.4 JOINT INFORMATION CENTER. .................................................... 71 7.3 . COMMUNICATIONS SYSTEMS AND NOTIFICATION ........................................ 72 7.3.1 PLANT COMMUNICATIONS EQUIPMENT .................................... 72 7.3.2 TELEPHONE COMMUNICATIONS ................................................. 72 7.3.3 SATELLITE TELEPHONES ............................................................. 73 7.3.4 RADIO COMMUNICATIONS ........................................................... 73 7.4 NOTIFICATION BY EMERGENCY CLASS .......................................................... 74 7.5 ENVIRONMENTAL ASSESSMENT CAPABILITIES ............................................. 75 7.5.1 SEISMIC MONITOR ........................................................................ 75 7.5.2 METEOROLOGICAL MONITORING ............................................... 75 7.5.3 MISSOURI RIVER MONITORING ................................................... 76 7.5.4 RADIOLOGICAL MONITORS ......................................................... 77 0 7.5.5 7.5.6 MAIN STEAM LINE MONITORS ............... :..................................... 78 ENVIRONMENTAL RADIATION SURVEILLANCE ......................... 80 7.5.7 RADIOANALYSIS LABORATORIES ............................................... 81 7.6 FIRE PROTECTION .............................................................................................81 7.7 EMERGENCY LOCKERS ..................................................................................... 81 7.8 HABITABILITY EQUIPMENT ................................................................................82 7.9 MEDICAL FACILITIES AND FIRST AID ............................................................... 83 7.9.1 MEDICAL FACILITIES .....................................................................83 7.9.2 FIRST AID ....................................................................................... 83

8. MAINTAINING EMERGENCY PREPAREDNESS ............................................................ 90 8.1 TRAINING ...............................................................*.............................................. 90 8.1.1 TRAINING FOR CNS EMERGENCY RESPONSE ORGANIZATION (ERO) .................................................................. 90 8.1.2 TRAINING FOR EMERGENCY PREPAREDNESS DEPARTMENT PERSONNEL ......................................................... 91 8.1.3 TRAINING FOR PARTICIPATING AGENCIES ............................... 91 8.1.4 PUBLIC EDUCATION ...................................................................... 92.

8.1.5 MEDIA FAMILIARIZATION .............................................................. 92 8.2 DRILLS AND EXERCISES ............................... ,.................................................... 93 8.2.1 EXERCISES ...,. ................................................................................93 8.2.2 DRILLS ............................................................................................95 8.3 EMERGENCY PREPAREDNESS DEPARTMENT ............................................... 96 8.4 CORPORATE COMMUNICATIONS DEPARTMENT ........................................... 97 8.5 REVIEW AND UPDATE OF THE EMERGENCY PLAN ....................................... 98

9. RECOVERY .....................................................................................................................99 EMERGENCY PLAN REVISION 76 PAGE 3 OF 192

9.1 RECOVERY PANEL .............................................................................................99 9.2 RECOVERY ORGANIZATION ............................................................................ 100 9.3 RECOVERY EXPOSURE CONTROL. ................................................................ 100 9.4 NUCLEAR SAFETY COMMITTEES ................................................................... 101 9.4.1 STATION OPERATIONS REVIEW COMMITTEE (SORG) ........... 101 9.4.2 SAFETY REVIEW AND AUDIT BOARD (SRAB) ........ ,.................. 101 APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN ....... 102 APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ........... 115 APPENDIX C EVACUATION ROUTES/MAPS .................................................... 166 APPENDIX D LETTERS OF AGREEMENT .......................................................... 175 APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES ................................ 178 APPENDIX F INTERFACING EMERGENCY PLANS .......................................... 192 0 EMERGENCY PLAN REVISION 76 PAGE 4 OF 192

LIST OF TABLES ') TABLE NO. TITLE PAGE 4.1-1 Notification of Unusual Event* 18 4.1-2 Alert* 21 4.1-3 Site Area Emergency* 24 4.1-4 General Emergency* 26 4.1-5 Notification of Unusual Event - Expected Actions 28 4.1-6 Alert - Expected Actions 29 4.1-7 Site Area Emergency - Expected Actions 30 4.1-8 General Emergency - Expected Actions 31 6.3-1 Assessment Actions 65 EPA Protective Action Guides (PAGs) for the Early Phase of a 6.4-1 66 Nuclear Incident 6.4-2 Initiation Times for Protective Actions for the General Public 67 7.1-1 ERF Communications Systems 89

CNS EAL Referenced Tables A-1, C-1, C-2, C-3, C-4, C-5, H-1, S-1, S-2, S-3, and F-1 along with the Notes Table can be found in the Emergency Plan Implementing Procedure 5.7.1, Emergency Classification.

EMERGENCY PLAN REVISION 76 PAGE 5 OF 192

LIST OF FIGURES

           \, FIGURE NO.                                 TITLE                           PAGE 5.2-1     CNS Nonnal Operating Organization - Control Room                47 CNS Emergency Response Organization - Technical Support 5.2-2                                                                     48 Center (TSC)

CNS Emergency Response Organization - Emergency 5.2-3 49 Operations Facility (EOF) CNS Emergency Response Organization - Operations Support 5.2-4 50 Center (OSC) CNS Emergency Response Organization - Joint lnfonnation 5.3-1 51 Center (JIC) 5.4-1 Interrelationships of Emergency Response Organizations 52 Federal Response Management Diagram - Cooper Nuclear 5.4-2 53 Station 7.2-1 TSC Floor Plan 84 7.2-2 OSC Floor Plan 85 7.2-3 EOF Floor Plan 86 7.2-4 JIC Floor Plan 87

O 7.4-1 Notification Chart for Emergency Classification 88 EMERGENCY PLAN REVISION 76 PAGE 6 OF 192

THE NEBRASKA PUBLIC POWER DISTRICT EMERGENCY PLAN FOR COOPER NUCLEAR STATION INTRODUCTION The Cooper Nuclear Station (CNS) is a one-unit Boiling Water Reactor power station rated at approximately 830 Mw(e) operated by the Nebraska Public Power District. The station is located on the west bank of the Missouri River between the towns of Brownville and Nemaha, Nebraska. Cooper Nuclear Station has been in commercial operation since July 1974. An independent spent fuel storage installation is located within the Protected Area of CNS. This Emergency Plan: (1) describes the organization formed to manage emergency situations; (2) provides the mechanism to classify emergencies according to severity of consequences; (3) defines and assigns functional responsibilities for emergency response actions; (4) outlines courses-of-action and protective measures to mitigate the consequences of an accident and to safeguard station personnel and the public; and (5) presents a general post-emergency plan and District organization for restoring the plant to normal operating status. Detailed implementing procedures for specific emergency actions during an incident are contained in the Cooper Nuclear Station Emergency Plan Implementing Procedures. This Emergency Plan establishes the policies and practices involving the Nebraska Public Power District in the unlikely event of an emergency at the Cooper Nuclear Station. Additions, deletions, or modifications to this Emergency Plan must be reviewed by the Station Operations Review Committee and approved by the Director of Nuclear Safety Assurance before such changes can be incorporated. The Director of Nuclear Safety Assurance has overall authority 0 and responsibility for radiological emergency response planning at CNS. The Emergency Plan Implementing Procedures for the Nebraska Public Power District emergency response activities are the documents which implement the requirements of this plan. Additions, deletions, or modifications to the Emergency Plan Implementing Procedures are made in accordance with the CNS Administrative Procedures. Copies of this Plan and the implementing documents shall be issued via a system of controlled distribution which will assure that all copies remain current. EMERGENCY PLAN REVISION 76 PAGE 7 OF 192

1. DEFINITIONS The following are definitions of tenns commonly used in the Nebraska Public Power District (NPPD) Emergency Plan.

1.1 Annual - Once during every calendar year. 1.2 Assessment Actions - Actions taken during or after an emergency to obtain and process information necessary to detennine the character and magnitude of the emergency and specific corrective emergency measures. 1.3 Class "A" Dose Assessment Model - A dose assessment computer code which utilizes actual 15 minute average meteorological data from the meteorological instrumentation maintained by the licensee or from alternate meteorological sources. This model provides calculations of relative concentrations and transit times within the plume exposure Emergency Planning Zone (EPZ). The output from a Class A model typically includes the plume dimensions and position, and the location, magnitude, and arrival time of (1) the peak relative concentration, and (2) the relative concentrations at appropriate locations. 1.4 Control Room - The Control Room, operating under the direction of the Shift Manager, is the primary point from which station conditions are monitored and controlled. It is the point where many corrective actions are taken to mitigate an emergency and where the initial assessment and classification of an emergency are made. 1.5 Corrective Actions - Measures tc;1ken to reduce the severity of, or-tenninate an 0 emergency at or near the source of the problem; to prevent an uncontrolled release of radioactive material; or to reduce the magnitude of the radioactive release. 1.6 Dosimeter of Legal Record (DLR) - A radiation dose monitoring device. A device used to determine an individual's accumulated external occupational radiation exposure including Deep Dose Equivalent (ODE), Lens Dose Equivalent (LOE), and Shallow Dose Equivalent (SOE). The term DLR is inclusive of Optically Stimulated Luminescent Dosimeters (OSLDs) and Thennoluminescent Dosimeters (TLDs).

1. 7 Emergency Action Levels - Parameter thresholds or sets of conditions used to classify an emergency. These parameters or conditions are indicators of the severity or potential severity of the emergency.

1.8 Emergency Operations Facility (EOF) - The Emergency Response Facility which is the focal point for overall NPPD Management of an emergency at CNS, and coordination of off-site radiologicar emergency operations. When activated, the EOF is under the direction of the EOF Director, who is responsible for maintaining continued coordination with governmental authorities regarding radiological consequences of an incident. 1.9 Emergency Planning Zone (EPZ) - Defined area established around CNS for which emergency planning is set forth in detail. These are the areas in which the potential need for protective action(s) is recognized and addressed. The two EPZs are defined as the plume exposure pathway and ingestion exposure pathway. EMERGENCY PLAN REVISION 76 PAGE 8 OF 192

1.10 Hostile Action - An act toward a nuclear power plant 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. 1.11 Information Authentication Center - That portion of the EOF where information concerning the emergency is gathered, coordinated, and disseminated. 1.12 Ingestion Exposure Pathway- The pathway through which principal exposure would be from the ingestion of contaminated water, milk, or food. The ingestion exposure pathway is referred to as the 50-mile EPZ since it includes the area within a 50-mile radius of CNS. 1.13 Joint Information Center (JIC) - The Off-Site Emergency Response Facility which is the prime location for coordinating news releases of information concerning the emergency between Utility, State, and Federal representatives. Employee information and rumor control activities are also coordinated from this location . 1.14 Legal Record - A document that satisfies State and Federal Regulations concerning radiation exposure to individuals. 1.15 Local Emergency Response Plans - Plans for local governmental response to radiological emergencies at CNS by Nemaha, Otoe, and Richardson counties in Nebraska and Atchison County in Missouri. Q 1.16 Operational Support Center (OSC) - The On-Site Emergency Response Facility from which Emergency Repair Teams, Monitoring Teams, Damage Control Teams, in-plant assignments, or other emergency response activities are coordinated and dispatched. The OSC is under the direction of the OSC Supervisor. 1.17 Plume Exposure Pathway - The pathway through which principal exposure is by whole body exposure to gamma radiation (from the plume and deposited materials) and inhalation exposure (from the passing radioactive plume). The time of potential exposure could range in length from minutes to days. The dimensions of the plume exposure planning zone are depicted in Appendix C and is also referred to as the 10-mile EPZ. 1.18 Process Radiation Monitoring System - Instrumentation designed to detect abnormal radiation levels and to activate appropriate alarms and controls for process and effluent plant system pathways. 1.19 Protective Actions - Actions taken to prevent or minimize radiological exposure. These may include in-house shelter, evacuation, respiratory protection, and thyroid blocking. 1.20 Protective Action Guides - The projected radiological dose which warrants protective action to minimize exposure to radioactive material. (

Reference:

"Manual of Protective Action Guides and Protective Actions for Nuclear Incidents" as revised May 1992, EPA 400-R-92-001.)

EMERGENCY PLAN REVISION 76 PAGE 9 OF 192

1.21 Recovery Actions - Post-emergency assessment, planning, resource allocation and corrective actions taken to restore the station as nearly as possible to its pre-emergency condition. 1.22 State Emergency Response Plans - Plans for the States of Nebraska, Missouri, Iowa, and Kansas responding to radiological emergencies at CNS. Each plan sets forth specific responsibilities and procedures for emergency agencies responsible for off-site emergency operations and the protection of the affected population. 1.23 Technical Support Center (TSC) - The On-Site Emergency Response Facility which provides space and equipment for emergency response personnel to monitor station conditions, analyze problems, and provide technical guidance and assistance to the Control Room, OSC, and EOF. It also contains technical documents and drawings, and is the focal point for on-site corrective action implementation during an emergency. This facility is under the direction of the TSC Director.

2. SCOPE AND APPLICABILITY 2.1 The Nebraska Public Power District (NPPD) Emergency Plan for CNS provides the mechanism to classify various types of emergencies and provides prior planning of emergency preparedness implementation actions. It delineates organized functions and responsibilities for the control and mitigation of an emergency to assure maximum protection of the public, station personnel, and plant equipment. This plan delineates responsibilities and actions to be taken by station, general office personnel, and other agencies during emergencies.

2.2 The Emergency Plan Implementing Procedures (EPIPs) designate responsibilities and define actions to be taken by assigned personnel in order to reduce or confine the consequences of an emergency. Appendix A provides the titles and summaries of the EPIPs. 2.3 The NPPD Emergency Plan interfaces with several State and Local Emergency Response Plans for areas that comprise the CNS Emergency Planning Zones (EPZs). Within the Nebraska portion of the 10-mile EPZ this Plan interfaces with the Nemaha County, Richardson County, and State of Nebraska Radiological Response Plans, as well as the Reception Area Plans for Otoe and Richardson Counties. Within the Missouri portion of the 10-mile EPZ, this plan interfaces with the Atchison County Nuclear Emergency Response Plan and the State of Missouri Nuclear Accident Plan. For the 50-mile Ingestion Pathway EPZ, the NPPD plan interfaces with the Radiological Response Plans of the states of Iowa and Kansas, as well as those of Nebraska and Missouri. In the event of a radiological emergency, the State and Local Agencies are responsible for coordinating their efforts in dealing with radiological concerns beyond the CNS site boundaries. 2.4 In addition to State and Local support, Federal agencies may also provide assistance in accordance with the National Response Framework (NRF) (i.e., Department of Homeland Security (OHS), Nuclear Regulatory Commission (NRG), Environmental Protection Agency (EPA), Department of Energy (DOE), Coast Guard, and Federal Emergency Management Agency (FEMA)). The NRG, acting as the coordinating agency, has technical leadership for the Federal government's response to an event. EMERGENCY PLAN REVISION 76 PAGE 1 0 OF 192

2.5 The protection of the health and safety of the general public is the prime concern; accordingly, the appropriate Local, State, and Federal Agencies will be supported by NPPD to the fullest extent practical.

3.

SUMMARY

OF THE NEBRASKA PUBLIC POWER DISTRICT (NPPD) CNS EMERGENCY PLAN 3.1 Nuclear power plant emergency plans are required to fulfill the requirements of 10CFR50.47 and Appendix E. The Nuclear Regulatory Commission (NRC) is required to determine the adequacy of the licensee Emergency Plan. The Federal Emergency Management Agency (FEMA) is required to detennine the adequacy of State and Local Plans. Together, the two agencies determine the adequacy of overall emergency preparedness. 3.2 The CNS Emergency Plan has been structured with NRC and FEMA guidance contained in NUREG-0654, Rev. 1 (FEMA-REP-1 ), "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants", Revision 1, November 1980. Appendix B of this plan provides a cross-index of the CNS Emergency Plan to NUREG-0654, Rev. 1 (FEMA REP 1). 3.3 The plan delineates the organization for emergencies, provides the methodology for classifying emergencies according to severity, defines and assigns responsibilities and authorities, and outlines measures to mitigate the consequences of an accident and minimize effects on the health and safety of the public and station personnel. In addition, the plan presents a general approach and organization for station recovery. 0 3.4 Radiological emergency planning for CNS has been coordinated with State and Local Emergency Response Agencies. The States of Nebraska, Missouri, Iowa, and Kansas, as well as the appropriate local government agencies which would be involved in emergency response operations, are aware of the emergency response measures described in the CNS Emergency Plan and will be advised of changes or modifications to these measures resulting from plan reviews and audits. 3.5 The CNS Emergency Response Organization (ERO) is responsible for on-site emergency operations and for maintaining a continuous flow of accurate radiological and station status infonnation to off-site emergency authorities. 3.6 Sections of the Emergency Plan in the balance of this document detail the emergency preparedness program. The contents of those sections are summarized below:

Section 4 - Emergency Conditions - Describes emergency classifications, initiating events, emergency actions levels, and corresponding NPPD and State and Local actions in response to each emergency classification. Emergency action levels and corresponding actions noted are based upon design and operating characteristics specific to CNS and described in NRC Endorsed Nuclear Energy Institute (NEI) Document 99-01, Revision 5, Methodology For Development of Emergency Action Levels.

EMERGENCY PLAN REVISION 76 PAGE 11 OF 192

Section 5 - Organizational Control of Emergencies - Describes the function and responsibilities of the CNS emergency response organization. Interfaces with off-site emergency agencies are defined and specified. This section also defines the specific assignments of personnel and identifies local and contract support service arrangements.
Section 6 - Emergency Measures - Describes the activation of the emergency response organization, assessment of plant conditions, initiation of protective and corrective actions on-site, recommendation of protective actions off-site, and measures to aid injured or contaminated personnel.
Section 7 - Emergency Response Facilities and Equipment - Describes facilities, emergency response equipment, and communications systems (on-site and off-site) available to assess emergency conditions, support emergency operations, notify off-site support agencies, provide aid to injured or contaminated personnel, and to control and mitigate incident-related damage.
Section 8 - Maintaining Emergency Preparedness - Describes the Emergency Preparedness Department and Emergency Preparedness Training Program, emergency drills and exercises, methods to review and update the Emergency Plan, and the process to maintain an adequate inventory of emergency equipment and supplies. This section also outlines methods used to provide emergency preparedness information to the general public.
Section 9 - Recovery - Defines, in general terms, post-emergency re-entry and recovery plans and organization. Recovery operations are divided into short-term activities, which are conducted during and immediately after an emergency, and long-term recovery activities, which comprise the more involved post-emergency efforts to return the station to a normal operating status.
Appendix A contains summaries of each EPIP and a cross-reference to the appropriate section of the CNS Emergency Plan.
Appendix B contains the cross-reference of the CNS Emergency Plan to NUREG-0654, Rev 1, (FEMA REP 1).
Appendix C contains maps and other references which depict evacuation routes, environmental sampling points, population distribution, etc., as defined in NUREG-0654, Rev. 1 Section J.10.(a) and (b)
Appendix D contains a listing of the Letters of Agreement maintained with off-site support agencies. The Emergency Plan signature page verifies that current Letters of Agreement are maintained in the Emergency Preparedness files as specified in Appendix D.
Appendix E contains a list of supplies and emergency equipment typical of the inventory kept in the emergency response equipment lockers and storage areas.
Appendix F contains a listing of the interfacing emergency response plans of the various Local, State, and Federal Support Agencies.

EMERGENCY PLAN REVISION 76 PAGE 12 OF 192

4. EMERGENCY CONDITIONS CNS maintains the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant Operators that an emergency action level has been exceeded and shall promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level.

The types of emergencies considered in the CNS Emergency Plan are classified into four categories as recommended in Emergency Action Level Guidelines for Nuclear Power Plants, Appendix 1, NUREG-0654, Rev. 1. The initiating conditions of NRG endorsed document NEI 99-01, Revision 5, *Methodology For Development of Emergency Action Levels", as well as the postulated accidents described in Chapter XIV of the Cooper Nuclear Station (CNS) Updated Safety Analysis Report, have been considered in developing the criteria presented in Section 4.1. Tables 4.1-1 through 4.1-4 provide specific Emergency Action Levels. Emergency Action Levels and corresponding classifications are included in EPIP 5. 7.1, Emergency Classification. Each successive classification is more severe. This classification system results in responses that are both timely and appropriate for a wide range of emergency conditions. There are three principal advantages of the graded classification system:

To assure timely notification of particular events which could lead to significant consequences should events continue to deteriorate, which might be indicative of more serious conditions not fully appreciated at the time of discovery.
To provide an assessment of the actual or likely implications of the event which can

() be clearly communicated to various affected parties during the early stages of the event.

To provide a means for setting in motion appropriate, prearranged, near-term emergency actions by affected parties.

4.1 CLASSIFICATIONS The four classifications of emergencies are:

NOTIFICATION OF UNUSUAL EVENT
ALERT
SITE AREA EMERGENCY
GENERALEMERGENC Y The fundamental logic connecting the four classifications of emergencies is to provide an escalating gradation of events related to the severity of their consequences.

Section 5 of the CNS Emergency Plan provides a description of the portions of the Emergency Response Organization which will be activated in the event of a NOTIFICATION OF UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY, or GENERAL EMERGENCY. EMERGENCY PLAN REVISION 76 PAGE 13 OF 192

4.1.1 NOTIFICATION OF UNUSUAL EVENT 4.1.1.1 Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs. These types of events may progress to a more severe emergency classification if they are not mitigated. Therefore, appropriate off-site agencies will be notified of such events in order to be better prepared for response if the event should progress to a more severe classification. 4.1.1.2 The purpose of this classification and its associated off-site notifications is to assure that the first step in any response later found to be necessary has been initiated. This brings the operating staff into a state of readiness, and provides a systematic handling of information and decision-making. These conditions may not be particularly significant from an emergency or safety standpoint, but have the potential to increase in significance from a safety standpoint if proper action is not taken or if circumstances beyond the control of the operating staff render the situation more serious. Upon declaration of a NOTIFICATION OF UNUSUAL EVENT, key on-site personnel, as well as specified management within NPPD will be notified (see Section 7.4). The NOTIFICATION OF UNUSUAL EVENT is maintained until the event is terminated or an escalation to a more severe emergency class is 0 required. 4.1.1.3 Table 4.1-1 lists Emergency Action Levels and Table 4.1-5 lists expected actions for the NOTIFICATION OF UNUSUAL EVENT classification. 4.1.2 ALERT 4.1.2.1 Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. At this classification, minor releases of radioactivity may occur or may have occurred. Operator modification of station operating status is a probable corrective action if such modification has not already been accomplished by automatic protection systems. 4.1.2.2 Upon declaration of an ALERT, notifications will be made (see Section 7.4 ). Notifying off-site agencies at an ALERT classification assures emergency personnel are readily available to respond if the situation becomes more serious, or to perform confirmatory radiation monitoring, if required. The TSC, EOF, and OSC are manned and activated at the declaration of an ALERT. EMERGENCY PLAN REVISION 76 PAGE 14 OF 192

4.1.2.3 The ALERT status is maintained until the event is downgraded, terminated, or escalated to a more severe emergency class. 4.1.2.4 Table 4.1-2 lists Emergency Action Levels and Table 4.1-6 lists expected actions for the ALERT classification. 4.1.3 SITE AREA EMERGENCY 4.1.3.1 Events are in process or have occurred which involve an actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. The SITE AREA EMERGENCY reflects conditions where there is a clear potential for significant releases of radioactive material, or such releases are in progress, but a core meltdown is not indicated based on current information. 4.1.3.2 Upon declaration of a SITE AREA EMERGENCY, non-ERO personnel are evacuated, monitoring teams are dispatched, off-site authorities are apprised of the emergency, the JIC is manned and activated, and periodic updates to the public are provided. 4.1.3.3 The SITE AREA EMERGENCY status is maintained until the event is 0 downgraded, terminated, or escalated to a GENERAL EMERGENCY. 4.1.3.4 Table 4.1-3 lists Emergency Action Levels and Table 4.1-7 lists expected actions for the SITE AREA EMERGENCY classification. 4.1.4 GENERAL EMERGENCY 4.1.4.1 Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off-site for more than the immediate site area. The GENERAL EMERGENCY reflects conditions that may affect the general public. The GENERAL EMERGENCY declaration initiates pre-determined protective actions for the public, provides information to the appropriate state, local, and federal authorities, initiates additional measures as indicated by actual or potential releases, provides for coordination with off-site authorities, and provides periodic updates for the public. EMERGENCY PLAN REVISION 76 PAGE 15 OF 192 I

4.1.4.2 Upon declaration of a GENERAL EMERGENCY, an automatic minimum protective action recommendation of evacuation for a 2-mile radius and 5 miles downwind, unless conditions make evacuation dangerous, and advise remainder of plume EPZ to go indoors to monitor EAS broadcasts will be made to state or local authorities. If conditions make evacuation dangerous, sheltering may be recommended as alternative protective action. Consider recommending evacuation of extended distances if conditions dictate. 4.1.4.3 The GENERAL EMERGENCY status is maintained until the event is downgraded or terminated. 4.1.4.4 Table 4.1-4 lists Emergency Action Levels and Table 4.1-8 lists expected actions for the GENERAL EMERGENCY classification. 4.2 OFF-SITE RADIOLOGICAL ASSESSMENT 4.2.1 The station will perform a preliminary assessment of the off-site consequences of an emergency. This preliminary assessment includes estimation by analytical methods of radiation dose rate, projected integrated dose for sectors and downwind distances, and a determination of the appropriate emergency classification. 4.2.2 The primary method for determining the radioactive release rate uses monitored release points. Effluent radiation monitor readings are available for CJ the Elevated Release Point, Turbine Building Vent, Reactor Building Vent, and Radwaste/Augmented Radwaste Building Vents. 4.2.3 The elevated release point release rate can be determined by correlating the exposure rates on high range radiation monitors in the drywell to those which have been calculated assuming a Design Basis Loss of Coolant Accident (LOCA). The LOCA calculations are based on the NUREG-0737 assumptions that of the maximum full power equilibrium isotopic inventories, 100% of the noble gases, 25% of the halogens, and 1% of the remaining particulates are instantaneously released to the atmosphere of the primary containment. The entire release is assumed to be through the Standby Gas Treatment System and out the elevated release point. Other methods may be used as described in EPIP 5.7.16. 4.2.4 The dose rate and integrated dose are based on duration of release, release rates, meteorological data, and atmospheric dispersion factors. 4.2.5 The radioiodine concentration is obtained by multiplying the radioiodine release rate by the dispersion factor. The Committed Dose Equivalent (COE) is determined by multiplying the air concentration by the exposure time and then by the dose conversion factor. 4.2.6 The noble gas concentration is obtained by multiplying the noble gas release rate by the dispersion factor. The Total Effective Dose Equivalent (TEDE) is determined by multiplying the air concentration by the exposure time and then by the appropriate dose conversion factor. EMERGENCY PLAN REVISION 76 PAGE 16 OF 192

4.2. 7 Upon activation of the EOF and the State Emergency Operations Centers, the affected state assumes primary responsibility for confirmatory and continuing off-site radiological assessment. This is accomplished by dispatching state Field Monitoring Teams and by analyzing data provided by the CNS Field Monitoring Teams. CNS will deploy Field Monitoring Teams for initial off-site monitoring prior to the arrival of responding State Field Monitoring Teams. These CNS teams may remain in the field to assist the state field monitoring teams. 4.3 SPECTRUM OF POSSIBLE ACCIDENTS AND INITIATING EVENTS 4.3.1 A number of accident scenarios which might occur at CNS have been analyzed in Chapter XIV of the CNS Updated Safety Analysis Report and within the Transnuclear NUHOMS Updated Final Safety Analysis Report for both severity of consequence and probability of occurrence. These types of accidents reflect the design characteristics of a Boiling Water Reactor and spent fuel storage installation and are addressed in Tables 4.1-1 through 4.1-4 and in EPIP 5.7.1 from the standpoint of initiating conditions, Emergency Action Levels, and subsequent emergency classification. EMERGENCY PLAN REVISION 76 PAGE 17 OF 192

TABLE 4.1-1 NOTIFICATION OF UNUSUAL EVENT ) EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE AU1.1 Any valid gaseous monitor reading > Table A-1 column "UE" ALL for~ 60 min. (Note 2). AU1.2 Any valid liquid effluent monitor reading > Table A-1 column ALL "UE" for~ 60 min. (Note 2). AU1.3 Confirmed sample analyses for gaseous or liquid releases ALL indicate concentrations or release rates > 2 x ODAM limits for

         ~ 60 min. (Note 2).

AU2.1 Unplanned water level drop in the reactor cavity or spent fuel ALL pool as indicated by any of the following:

Ll-86 (calibrated to 1001' elev.)
Spent fuel pool low level alarm
Visual observation AND
Valid area radiation monitor reading rise on RMA-RA-1 or RMA-RA-2.

0 AU2.2 Unplanned valid area radiation monitor reading or survey results rise by a factor of 1,000 over normal levels*. ALL

         *Normal levels can be considered as the highest reading in the past 24 hours excluding the current peak values.

CU1.1 AC power capability to critical 4160V Buses 1F and 1G MODES 4 or reduced to a single power source (Table C-4) for~ 15 min. 5 such that any additional single failure would result in loss of all AC power to critical buses (Note 3). CU2.1 RPV level cannot be restored and maintained> +3 in. for MODE4

         ~ 15 min. (Note 3) due to RCS leakage.

CU2.2 Unplanned RPV level drop for~ 15 min. (Note 3) below, MODE5 EITHER: RPV flange (Ll-86: 206 in. normal calibration, 113.75 in. elevated calibration) OR RPV level band when the RPV level band is established below the RPV flange. CU2.3 RPV level cannot be monitored with any unexplained RPV MODE 5 leakage indication, Table C-1. CU3.1 Any unplanned event results in RCS temperature > 212°F due MODES 4 or to loss of decay heat removal capability. 5 CU3.2 Loss of all RCS temperature and RPV level indication for MODES 4 or

         ~ 15 min. (Note 3).                                                 5 EMERGENCY PLAN                                    REVISION  76              PAGE 18 OF 192

TABLE 4.1-1 NOTIFICATION OF UNUSUAL EVENT EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE CU4.1 Loss of all Table C-2 on-site (internal) communication methods MODES 4, 5, affecting the ability to perform routine operations OR Loss of or all Table C-2 off-site (external) communication methods DEFUELED affecting the ability to perform off-site notifications. CU5.1 An unplanned sustained positive period observed on nuclear MODES 4or instrumentation. 5 CU6.1 < 105 VDC bus voltage indications on all Technical MODES 4 or Specification required 125 VDC buses for~ 15 min. (Note 3). 5 FU1.1 Any loss or any potential loss of Primary Containment. MODES 1, 2, (Table F-1) or3 HU1.1 Seismic event identified by any two of the following: ALL

The Seismic Monitor System free field sensor actuated or Alarm B-3/B-1 seismic event.
Earthquake felt in plant.
National Earthquake Information Center.
O

\ ____ HU1.2 Tornado striking within Protected Area boundary OR sustained high winds~ 100 mph. ALL HU1.3 Main turbine failure resulting in casing penetration or damage ALL to turbine or generator seals. HU1.4 Flooding in any Table H-1 area that has the potential to affect ALL safety-related equipment required by Technical Specifications for the current operating mode. HU1.5 High river/forebay water level > 899' MSL OR low river ALL level/forebay < 870' MSL. HU2.1 Fire in any Table H-1 area not extinguished within 15 min. of ALL Control Room notification or receipt of a valid Control Room alarm due to fire (Note 3). HU2.2 Explosion within the Protected Area. ALL HU3.1 Toxic, corrosive, asphyxiant, or flammable gases in amounts ALL that have or could affect normal plant operations. HU3.2 Recommendation by local, county, or state officials to evacuate ALL or shelter site personnel based on an off-site event. HU4.1 A security condition that does not involve a hostile action as ALL reported by the Security Shift Supervisor OR a credible site-specific security threat notification OR a validated notification from NRG providing information of an aircraft threat. EMERGENCY PLAN REVISION 76 PAGE 19 OF 192 I

TABLE 4.1-1 NOTIFICATION OF UNUSUAL EVENT EMERGENCY ACTION LEVELS (EALs)

OPERATING EAL (Alarm, Instrument Reading, etc.) MODE HU6.1 Other conditions exist which in the judgment of the Emergency ALL Director indicate that EITHER: Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant OR a security threat to facility protection has been initiated. No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs. SU1.1 Loss of all off-site AC power (Table S-3) to critical 4160V MODES 1, 2, Buses 1F and 1G for ~ 15 min. (Note 3). or3 SU2.1 An unplanned sustained positive period observed on nuclear MODE3 instrumentation. SU3.1 Plant is not brought to required operating mode within MODES 1, 2, Techni~I Specifications.LCO action statement time. or3 SU4.1 Unplanned loss of> approximately 75% of annunciators or MODES 1, 2, 0 '\.. indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C for~ 15 min. (Note 3). or3 SU5.1 MODES 1, 2, SJAE monitor> 1.58E+3 mR/hr. or3 SU5.2 MODES 1, 2, Coolant activity~ 4.0 µCi/gm dose equivalent 1-131. or3 SU6.1 Unidentified or pressure boundary leakage > 10 gpm OR MODES 1, 2, Identified leakage > 30 gpm (Note 6). or3 SU8.1 Loss of all Table S-2 on-site (internal) communications MODES 1, 2, capability affecting the ability to perform routine operations or3 OR loss of all Table S-2 off-site (external) communications methods affecting the ability to perform off-site notifications. EU1.1 Damage to a loaded cask confinement boundary. N/A EMERGENCY PLAN REVISION 76 PAGE 20 OF 192 I

TABLE 4.1-2 ALERT EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE AA1.1 Any valid gaseous monitor reading> Table A-1 column "Alert" ALL for~ 15 min. (Note 2). AA1.2 Any valid liquid effluent monitor reading > Table A-1 column ALL "Alert" for~ 15 min. (Note 2). AA1.3 Confirmed sample analyses for gaseous or liquid releases ALL indicate concentrations or release rates > 200 x ODAM limits for~ 15 min. (Note 2). AA2.1 Damage to irradiated fuel OR loss of water level (uncovering ALL irradiated fuel outside the RPV) that causes EITHER of the following: Valid RMA-RA-1 Fuel Pool Area Rad reading> 5.0E+04 mR/hr OR valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum Hi-Hi alarm. AA2.2 A water level drop in the reactor refueling cavity or spent fuel ALL pool that will result in irradiated fuel becoming uncovered. 0 AA3.1 Dose rates > 15 mRem/hr in EITHER of the following areas ALL requiring continuous occupancy to maintain plant safety functions: Main Control Room (RM-RA-20) OR GAS. CA1.1 MODES 4, 5, Loss of all off-site and all on-site AC power (Table C-4) to or critical 4160V Buses 1 F and 1G for > 15 min. (Note 3). DEFUELED CA2.1 RPV level < -42 in. OR RPV level cannot be monitored for MODES 4 or

           ~ 15 min. (Note 3) with any unexplained RPV leakage              5 indication, Table C-1.

CA3.1 Any unplanned event results in EITHER: MODES 4 or 5 RCS temperature> 212°F for> Table C-3 duration (Note 4) OR RPV pressure increase > 10 psig due to a loss of RCS cooling. FA1.1 Any i'oss or any potential loss of either Fuel Clad or RCS MODES 1, 2, (Table F-1 ). or3 I EMERGENCY PLAN REVISION 76 PAGE 21 OF 192 I

TABLE 4.1-2 ALERT EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alam,, Instrument Reading, etc.) MODE HA1.1 Seismic event > 0.1 g as indicated by the Seismic Monitor ALL System free field sensor or Alam, B-3/A-1, EMERGENCY SEISMIC HIGH LEVEL, AND earthquake confimied by any of the following:

Earthquake felt in plant.
National Earthquake lnfomiation Center.
Control Room indication of degraded performance of systems required for the safe shutdown of the plant.

HA1.2 Tornado striking or high winds~ 100 mph resulting in EITHER: ALL Visible damage to any Table H-1 area structure containing safety systems or components OR Control Room indication of degraded performance of safety systems. HA1.3 Main turbine failure-generated projectiles result in EITHER: ALL Visible damage to or penetration of any Table H-1 area structure containing safety systems or components OR Control 8 Room indication of degraded perfomiance of safety systems. ALL HA1.4 Flooding in any Table H-1 area resulting in EITHER: An electrical shock hazard that precludes access to operate or monitor safety equipment OR Control Room indication of degraded perfomiance of safety systems. HA1.5 High river/forebay water level > 902' MSL OR low river/forebay ALL level < 865' MSL. HA1.6 Vehicle crash resulting in EITHER: ALL Visible damage to any Table H-1 area structure containing safety systems or components OR Control Room indication of degraded performance of safety systems. HA2.1 Fire or explosion resulting in EITHER: ALL Visible damage to any Table H-1 area containing safety systems or components OR Control Room indication of degraded performance of safety systems. HA3.1 Access to any Table H-1 area is prohibited due to toxic, ALL corrosive, asphyxiant, or flammable gases which jeopardize operation of systems required to maintain safe operations or safely shutdown the reactor (Note 7). EMERGENCY PLAN REVISION 76 PAGE 22 OF 192

TABLE 4.1-2 ALERT EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE HA4.1 A hostile action is occurring or has occurred within the Owner ALL Controlled Area as reported by the Security Shift Supervisor OR a validated notification from NRC of an airliner attack threat wtthin 30 min. of the site. HA5.1 Procedure 5.1ASD, Alternate Shutdown, or ALL Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside the Control Room, requires Control Room evacuation. HA6.1 Other conditions exist which in the judgment of the Emergency ALL Director indicate that events are in progress or have occurred which involve EITHER: 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 0 the EPA Protective Action Guideline exposure levels beyond the site boundary. SA1.1 AC power capability to critical 4160V Buses 1F and 1G MODES 1, 2, reduced to a single power source (Table S-3) for> 15 min. or 3 such that any additional single failure would result in loss of all AC power to critical buses (Note 3). SA2.1 An automatic scram failed to shut down the reactor AND MODES 1 or manual actions taken at the reactor control console (Note 5) 2 successfully shut down the reactor as indicated by reactor power< 3%. SA4.1 Unplanned loss of> approximately 75% of annunciators or MODES 1, 2, indicators associated with safety systems on Control Room or3 Panels 9-3, 9-4, 9-5, and C for :2: 15 min. (Note 3) AND EITHER: Any significant transient is in progress, Table S-1 OR Compensatory indications are unavailable. EMERGENCY PLAN REVISION 76 PAGE 23 OF 192

TABLE 4.1-3 SITE AREA EMERGENCY '.) EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE AS1.1 Any valid gaseous monitor reading> Table A-1 column "SAE" ALL for~ 15 min. (Note 1). AS1.2 Dose assessment using actual meteorology indicates doses ALL

           > 0.1 Rem lEDE or> 0.5 Rem thyroid COE at or beyond the site boundary.

AS1.3 Field survey indicates closed window dose rates > 0.1 Rem/hr ALL that is expected to continue for::::: 60 min. at or beyond the site boundary (Note 1) OR field survey sample analysis indicates thyroid COE > 0.5 Rem for 1 hr of inhalation at or beyond the site boundary. CS2.1 With Containment Closure not established, RPV level < -48 in. MODES 4 or (Note 4). 5 CS2.2 With Containment closure established (Note 4 ), RPV level MODES 4 or

           < -158 in. (Note 4).                                                5 CS2.3  RPV level cannot be monitored for::::: 30 min. (Note 3) with a      MODES 4 or 0           loss of inventory as indicated by EITHER:

Unexplained RPV leakage indication, Table C-1, OR Erratic 5 Source Range Monitor indication. FS1.1 Loss or potential loss .of any two barriers (Table F-1 ). MODES 1, 2, or3 HS4.1 A hostile action is occurring or has occurred within the ALL Protected Area as reported by the Security Shift Supervisor. HS5.1 Control Room evacuation has been initiated AND control of the ALL plant cannot be established within 15 min. EMERGENCY PLAN REVISION 76 PAGE 24 OF 192 I

~*

TABLE 4.1-3 SITE AREA EMERGENCY EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE HS6.1 Other conditions exist which in the judgment of the Emergency ALL Director indicate that events are in progress or have occurred which involve EITHER: An actual or likely major failures of plant functions needed for protection of the public OR Hostile action that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels (1 Rem TEDE and 5 Rem thyroid COE) beyond the site boundary. SS1.1 Loss of all off-site and all on-site AC power (Table S-3) to MODES 1, 2, critical 4160V buses 1F and 1G for:?: 15 min. (Note 3). or3 10 SS2.1 An automatic scram failed to shut down the reactor AND manual actions taken at the reactor control console (Note 5) MODES 1 or 2 do not shut down the reactor as indicated by reactor power:?: 3%. SS4.1 Loss of> approximately 75% of the annunciators or indicators MODES 1, 2, associated with safety systems on Control Room Panels 9-3, 9- or3 4, 9-5, and C for z 15 min. (Note 3) AND any significant transient is in-progress, Table S-1 AND compensatory indications are unavailable. SS7.1 < 105 VDC bus voltage indications on all vital 125 VDC buses MODES 1, 2, (1A and 18) for z 15 min. (Note 3). or3 EMERGENCY PLAN REVISION 76 PAGE 25 OF 192

TABLE 4.1-4 GENERAL EMERGENCY EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE AG1 .1 Any valid gaseous monitor reading > Table A-1 column "GE" for ALL 2 15 min. (Note 1). AG1 .2 Dose assessment using actual meteorology indicates doses ALL

          > 1 Rem TEDE or > 5 Rem thyroid COE at or beyond the site boundary.

AG1 .3 Field survey results indicate closed window dose rates ALL

          > 1 Rem/hr expected to continue for z 60 min. at or beyond the site boundary (Note 1) OR analyses of field survey samples indicate thyroid COE > 5 Rem for 1 hr of inhalation at or beyond the site boundary.

CG2.1 RPV level < -158 in. for~ 30 min. (Note 3) AND any MODES 4 or 5 Containment Challenge indication, Table C-5. CG2.2 RPV level cannot be monitored for z 30 min. (Note 3) with core MODES 4 or 5 uncovery indicated by EITHER: Unexplained RPV leakage indication, Table C-1 OR Erratic () Source Range Monitor indication AND any Containment Challenge indication, Table C-5. FG1 .1 Loss of any two barriers AND loss or potential loss of third MODES 1, 2, barrier (Table F-1 ). or3 HG4.1 A hostile action has occurred such that plant personnel are ALL unable to operate equipment required to maintain safety functions OR a hostile action has caused failure of Spent Fuel Cooling Systems and imminent fuel damage is likeJy for a freshly off-loaded reactor core in pool. HG6.1 Other conditions exist which in the judgment of the Emergency ALL Director indicate that events are in progress or have occurred which involve EITHER: 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 (1 Rem TEDE and 5 Rem thyroid COE) beyond the site boundary. EMERGENCY PLAN REVISION 76 PAGE 26 OF 192

TABLE 4.1-4 GENERAL EMERGENCY EMERGENCY ACTION LEVELS (EALs) OPERATING EAL (Alarm, Instrument Reading, etc.) MODE SG1 .1 Loss of all off-site and all on-stte AC power (Table S-3) to MODES 1, 2, critical 4160V buses 1F and 1G AND EITHER: or3 Restoration of at least one emergency bus in < 4 hours is not likely OR RPV level cannot be restored and maintained > -158 in. or cannot be determined. SG2.1 Automatic and all manual scrams were not successful AND Modes 1 or 2 Reactor power z 3% AND EITHER of the following exist or have occurred due to continued power generation: RPV level cannot be restored and maintained> -183 in. or cannot be determined OR Average torus water temperature and RPV pressure cannot be maintained within the Heat Capacity Temperature Limtt (EOP/SAG Graph 7). 0 EMERGENCY PLAN REVISION 76 PAGE 27 OF 192

TABLE 4.1-5 NOTIFICATION OF UNUSUAL EVENT- EXPECTED ACTIONS CNS ACTIONS STATE/LOCAL ACTIONS

1. Notrfy responsible State and Local 1. Provide assistance if requested (fire, Governmental Agencies of emergency security, medical, etc.).

conditions within 15 minutes of declaration. 2. Continue notification as necessitated by situation.

2. Notrfy the ERO if determined to be necessary by the Emergency Director. 3. Standby until verbal termination.

OR

3. Augment on-shift resources as needed.

4. Escalate to a more severe class.

4. Assess and respond.

5. Terminate with verbal summary to off-site authorities, followed by written report within 24 hours.

OR

6. Escalate to a more severe class.

0 EMERGENCY PLAN REVISION 76 PAGE 28 OF 192

TABLE 4.1-6 ALERT - EXPECTED ACTIONS CNS ACTIONS STATE/LOCAL ACTIONS

1. Notify responsible State and Local 1. Provide assistance tf requested (fire, Governmental Agencies of emergency security, medical, etc.).

conditions within 15 minutes of declaration. 2. Augment resources and bring primary response centers to stand-by stcitus.

2. Notify the ERO, augment resources and activate TSC, OSC, and EOF. The JIC 3. Place key emergency personnel may be placed on standby status. (including monitoring teams and associated communications) on

3. Assess and respond. stand-by status.

4. Dispatch on-site monitoring teams with 4. If necessary, Governor proclaims associated communications. Disaster Emergency Condition.

Dispatch State Field Command Post

5. Provide periodic plant status updates to and key emergency response off-site authorities. personnel including Radiological Monitoring Teams with associated

6. Provide periodic meteorological equipment and communications. Alert assessments to off-site authorities and, if all state agencies and local government

O releases .are occurring, dose estimates for actual releases.

to standby or assume an increased readiness posture.

7. Close out or recommend reduction in 5. Provide confirmatory off-site radiation Emergency Class by verbal summary to monitoring and ingestion pathway dose off-site authorities followed by written projections if actual releases summary within 8 hours of closeout or substantially exceed Off-Site Dose class reduction. Assessment Manual (ODAM) limits.

OR

6. Maintain ALERT status until verbal

8. Escalate to a more severe class. termination.

OR

7. Escalate to a more severe class.

EMERGENCY PLAN REVISION 76 PAGE 29 OF 192

., 1. CNS ACTIONS Notify responsible State and Local TABLE 4.1-7 SITE AREA EMERGENCY - EXPECTED ACTIONS 1. STATE/LOCAL ACTIONS Provide assistance as requested in Governmental Agencies of emergency accordance with established disaster conditions within 15 minutes of support procedures. declaration.

2. If In-House Shelter is desirable,

2. Notify the ERO, augment resources by activate public notification systems activating TSC, OSC, EOF and JIG. within at least two miles of the plant.

3. Assess and respond. 3. Provide the public within the plume exposure EPZ with periodic updates

4. Dispatch on-site and off-site monitoring on emergency status.

teams with communications equipment.

4. Augment resources by activating

5. Provide a dedicated individual for plant state/local EOCs.

status updates to off-site authorities and periodic press briefings. 5. Dispatch key on-site emergency personnel, including monitoring tea111s

6. Make Senior Technical and and communications equipment.

Management staff on-site available for consultation with NRG and state 6. Alert other personnel to standby representatives on a periodic basis. status (e.g., those needed for traffic 0 7. Provide meteorological data and dose control or evacuation) and dispatch personnel to near-site duty station. estimates (for actual releases) to off-site authorities via a designated 7. Provide off-site monitoring results to individual. CNS and jointly assess them.

8. Provide release and dose projections 8. Continuously assess information from based on available plant condition CNS and off-site monitoring teams information or contingencies. with regard to initiating/modifying public protective actions.

9. Closeout or recommend reduction in emergency class by briefing off-site 9. Recommend placing milk animals authorities at EOF followed by written within two miles on stored feed and summary within 8 hours of closeout or assess need to extend distance.

class reduction.

10. Provide press briefings, perhaps OR jointly with CNS.

10. Escalate to GENERAL EMERGENCY.

11. Maintain SITE AREA EMERGENCY status until termination or reduction of emergency class.

OR ,_) 12. Escalate to GENERAL EMERGENCY EMERGENCY PLAN REVISION 76 PAGE 30 OF 192

TABLE 4.1-8 GENERAL EMERGENCY - EXPECTED ACTIONS

  ;---)                                                    STATE/LOCAL ACTIONS CNS ACTIONS

1. Notify responsible State and Local 1. Provide any assistance requested in Governmental Agencies of emergency accordance with established disaster conditions within 15 minutes of support procedures.

declaration. 2. Initiate immediate public notification

2. Notify the ERO, augment resources by of GENERAL EMERGENCY status activating TSC, OSC, EOF, and JIG. and provide periodic public updates.

3. Recommend protective action of 3. Recommend evacuation for a 2-mile evacuation for a 2-mile radius and radius and a 5-mile direct downwind 5 miles downwind, unless conditions segment and assess the need to make evacuation dangerous, and extend distances if conditions advise remainder of plume EPZ to go degrade.

indoors to monitor EAS broadcasts to 4. Augment resources by activating State and Local Authorities. Consider state/local primary response centers. recommending evacuation of extended distances if degrading conditions 5. Dispatch other emergency personnel warrant. to duty stations within 5-mile radius and alert others to standby status.

4. Assess and respond.

6. Provide off-site monitoring results to

5. Dispatch on-site and off-site monitoring

..~) teams and associated communications equipment. 7. CNS, DOE, EPA, and others and jointly assess them . Continuously assess information from

6. Provide a dedicated individual for plant CNS and off-site monitoring teams status updates to off-site authorities with regard to modifying public and periodic press briefings. protective actions and mobilizing Coordinate joint information releases evacuation resources.

with off-site authorities.

8. Recommend placing milk animals

7. Make Senior Technical and within 10-mile radius on stored feed Management Staff available for and assess need to extend distance.

periodic consultation with NRG and state representatives. 9. Provide press briefings, perhaps jointly with CNS.

8. Provide meteorological data and dose estimates (for actual releases) to 10. Maintain GENERAL EMERGENCY off-site authorities via a dedicated status until termination or reduction of individual. emergency class.

9. Provide release and dose projections based upon available plant condition information and foreseeable contingencies.

10. Tem1inate (or recommend reduction of) emergency class by briefing off-site authorities at the EOF, followed by written summary within 8 hours.

EMERGENCY PLAN REVISION 76 PAGE 31 OF 192

5. ORGANIZATIONAL CONTROL OF EMERGENCIES In the event of an emergency, NPPD has both the normal operating organization (on-shift Control Room staff) and an organization specifically designed to augment them. The Emergency Response Organization may, depending upon the classification of the accident, range from the normal operating staff to a comprehensive emergency force composed of on-site, general office, state, and local support, and contract personnel.

This section of the Emergency Plan describes the normal on-shift Operating organization, the Emergency Response Organization, other support available, and the governmental agencies responsible for dealing with off-site emergency conditions. Post-emergency station recovery plans are presented in Section 9. A general overview of the Nuclear Power Group Organization is described in the USAR. 5.1 NORMAL OPERATING ORGANIZATION Direct operation and control of the Nuclear Power Plant is the responsibility of the on-duty Operations Crew operating out of the Control Room. The normal operating crew is staffed and qualified to perform all actions necessary to institute immediate protective measures and to implement the Emergency Plan. The composition and relationship of the Control Room crew (Normal Operating Organization) is depicted in Figure 5.2-1. 5.1.1 LINES OF AUTHORITY 5.1.1.1 The Shift Manager is the senior licensed individual on the operating crew. The Shift Manager must hold a Senior Reactor Operator's license. 5.1.1.2 If the Shift Manager is incapacitated the Control Room Supervisor assumes the Shift Manager role. The Control Room Supervisor must also hold a Senior Reactor Operator's license. 5.1.1.3 Reactor Operators (Control Room Operators) report to the Control Room Supervisor. Reactor Operators must hold a Reactor Operator's license 5.1.1.4 Station Operators and Utility/Fire Brigade personnel function under the direction of Reactor Operators or the Control Room Supervisor. Station Operators and Utility/Fire Brigade do not require a license. 5.1.1.5 The Chem/RP Technician reports to the Shift Manager. 5.1.1.6 The Shift Technical Engineer reports to the Shift Manager. 5.1.1. 7 The Dose Assessor reports to the Shift Manager. 5.1.1.8 The Shift Communicator reports to the Shift Manager. EMERGENCY PLAN REVISION 76 PAGE 32 OF 192

5.1.2 RESPONSIBILITIES/FUNCTIONS 5.1.2.1 The Shift Manager is the individual, on-shift at all times, vested with the authority and responsibility to immediately and unilaterally initiate any emergency actions, including protective action recommendations to authorities responsible for implementing off-site emergency measures. Upon declaration of an emergency, the Shift Manager becomes the site Emergency Director. The Emergency Director is responsible for, and may not delegate, classifying emergencies, ensuring notifications are made to off-site authorities, and recommending protective actions to off-site authorities. The Shift Manager is responsible for providing guidance and support to the operating crew. He will ensure that there is an adequate staff to perform the required operational functions and be responsible for ensuring proper communications from the Control Room to the TSC, OSC, and EOF. In conjunction with key technical personnel, he will also assess station operations and ensure recommended corrective actions are given adequate consideration. He will also control and monitor station conditions, take corrective actions to mitigate or terminate the incident, stabilize the plant, and minimize accident consequences. Upon entry into Severe Accident *Guidelines, the Shift Manager shall take direction for accident mitigation from the Operations Coordinator in the TSC. () Upon activation of the EOF, the on-call Emergency Director wi,11 relieve the Shift Manager of Emergency Director duties. 5.1.2.2 The Control Room Supervisor (CRS) directs the activities of the Control Room Operators and Station Operators in response to normal, abnormal, and emergency procedures. The CRS functions as the primary command and control interface between the Shift Manager and other Operations Shift personnel. 5.1.2.3 Reactor Operators (Control Room Operators) are responsible for the safe operation of the reactor and balance of plant. 5.1.2.4 Station Operators are responsible to the Control Room Supervisor. Station Operators perform functions in the plant such as equipment monitoring, log keeping, equipment operation, and tag-outs. 5.1.2.5 The Chem/RP Technician is responsible for providing advice and assistance to the crew regarding radiological issues. 5.1.2.6 The Shift Communicator is responsible for all off-site communication duties. 5.1.2. 7 The Dose Assessor has the primary responsibility for performing Dose Assessment on-shift. EMERGENCY PLAN REVISION 76 PAGE 33 OF 192

5.1.2.8 The Shift Technical Engineer serves in an advisory capacity to the Shift Manager in the diagnosis of off-nonnal events, transients, and accident situations. The Shift Technical Engineer functions to provide an independent assessment of the operation and response of the plant. The STE may perfonn other functions as assigned, so long as they do not interfere with this oversight role. Upon declaration of an emergency (Alert or higher), two additional positions become available to the Control Room. They are part of the ERO (described below) but function from the Control Room, therefore are described here. 5.1.2.9 The Technical Communicator provides a flow of technical data to the OPS/EOP Advisors in the TSC and EOF, and the Technical Communicator in the OSC. 5.1.2.10 The Control Room Log keeper maintains an accurate log of important Control Room activities. 5.2 EMERGENCY RESPONSE ORGANIZATION Key elements of the Emergency Response Organization (ERO) at CNS are depicted in Figures 5.2-2, 5.2-3, and 5.2-4. Emergency Response Organization (ERO) positions in the TSC, OSC, and EOF, along with functions for these facilities are summarized in the following paragraphs. 0 Sufficient personnel have been designated to assure that functional responsibilities are maintained for continuous 24-hour operation. The Emergency Response Organization roster is maintained by the Emergency Preparedness Department. The Emergency Response Organization operates from the Control Room and the following Emergency Response Facilities. These CNS emergency Response Facilities will be activated within approximately one (1) hour following the declaration of an Alert or higher.

Technical Support Center.
Operational Support Center or alternate.
Emergency Operations Facility.

The Emergency Response Organization is supported by the following Emergency Response Facility as necessary:

Joint lnfonnation Center.

The TSC, EOF, and OSC will be manned and activated at the declaration of an ALERT or higher level emergency. The JIC may be placed on standby during an ALERT, and will be manned and activated upon the declaration of a SITE AREA EMERGENCY or GENERAL EMERGENCY. EMERGENCY PLAN REVISION 76 PAGE 34 OF 192

In all emergency classifications, the Emergency Director is in charge of the Emergency Response Organization. The Emergency Director is assigned the authority and responsibility to immediately and unilaterally initiate emergency response actions. The Emergency Director may not delegate the following:

Event Declaration.
The decision to notify authorities responsible for off-site emergency measures.
Recommend protective actions to authorities responsible for off-site emergency measures.

Under a NOTIFICATION OF UNUSUAL EVENT, all emergency response functions will usually be conducted from the Control Room by the on-shift Operating organization described in Section 5.1. At an ALERT, the TSC, EOF, and OSC will be activated and will provide further management, technical, and craft support. At a SITE AREA EMERGENCY or GENERAL EMERGENCY, the JIC is activated and will have the additional support of the JIC Director. The authorities and responsibilities of each position are as follows: 5.2.1 EMERGENCY DIRECTOR The Emergency Director is in command of the NPPD Emergency Response Organization. His/her responsibilities are as follows:

Verify that the NPPD on-site and off-site emergency response functions are being performed in a timely manner.

0

Ensure that adequate technical and logistical support is available to the station organization.
Ensure continuity of emergency response resources.

All emergency actions which may involve exposures exceeding occupational exposure limits must be approved by the Emergency Director. The Emergency Director provides management expertise to the emergency organization and may initially report to the Control Room instead of the EOF. As the situation warrants, he may relocate to any on-site facility to confer with members of the various emergency response organizations. The Emergency Director will be supported by the following positions: EMERGENCY PLAN REVISION 76 PAGE 35 OF 192

5.2.2 TSC DIRECTOR The TSC Director is in charge of TSC functions and activities. His/her primary responsibility is to maintain command and control in the TSC to provide technical assistance and recommendations to the Control Room. The primary function of the TSC staff is to augment Control Room efforts to manage the plant emergency by:

Diagnosing station conditions.
Recommending and prioritizing corrective or mitigative actions.
Providing technical support to Control Room personnel.

The TSC Director is assisted in these functions by individuals assuming the minimum staff ERO positions below and depicted in Figure 5.2-2. 5.2.2.1 The Operations Coordinator provides a liaison between the Control Room and the TSC/OSC Staffs, on personnel, technical, and administrative issues related to plant operations. Upon entry into Severe Accident Guidelines, the Operations Coordinator shall assume decision-making authority from the Shift Manager related to accident mitigation actions and provide direction to the Control Room operating staff. 5.2.2.2 The Engineering Coordinator provides engineering expertise to the TSC Director. He/she shall also coordinate the activities of the Engineering Group through the Engineering Team Leader. The Engineering Coordinator will maintain liaison with General Electric, POWER Engineers (formerly Bums & Roe, Inc.), Institute of Nuclear Power Operations, and other contract support as referenced in Section 5.3.3. 5.2.2.3 The Maintenance Coordinator provides expertise to the TSC Director in the areas of equipment analysis/status, repair options, and equipment repair priorities. The Maintenance Coordinator also supervises the activities of the OSC as directed by TSC Director through the OSC Supervisor. 5.2.2.4 The Chemistry/Radiological Protection Coordinator provides chemistry and radiological protection expertise to the TSC Director and is also responsible for the following:

Assess radiological dose, recommend radiation protection measures, direct radiological surveys and decontamination actions, and assist in assessment of off-site consequences.
Provide chemical analyses for the evaluation of station systems and provide data to aid in the determination of reactor core conditions and release potentials.

EMERGENCY PLAN REVISION 76 PAGE 36 OF 192

This individual is assisted by other radiological personnel. These emergency response personnel will provide technical expertise on radiological release rates and dose projections, in plant radiological surveys, and will input data into the dose assessment model, when required.

NOTE - The additional Key Functional staff, and other positions listed below, enhance the operation of the TSC. 5.2.2.5 The ENS Communicator will provide continuous communication with the NRG when requested to do so. 5.2.2.6 The Operations/EOP Advisor provides technical assistance and operational information to the Operations Coordinator. He/she maintains a proactive assessment of EOP and SAG implementation as well as performs plant condition assessments. 5.2.2. 7 The Engineering Team Leader directs the efforts of the Engineering group based on the direction and priorities established by the TSC Director and Engineering Coordinator. 5.2.2.8 The Electrical Engineer provides information on station electrical system capabilities, status, alternate power arrangements, and evaluates the necessity of repair, installation, and modification of electrical equipment. The Electrical Engineer will also provide information on l&C issues. 5.2.2.9 The Mechanical Engineer performs analyses on mechanical components and provides information on various mechanical systems capabilities, status, and evaluates the necessity of repair, installation, or modification of mechanical equipment. 5.2.2.10 The Civil Engineer provides information and analysis on station component structural status and integrity. 5.2.2.11 The Reactor Engineer provides information and analysis on the conditions of the reactor core. 5.2.2.12 The Function Status Assessment Engineer evaluates the availability of plant systems which may be used to perform functions specified in the Plant Specific Technical Guidelines/Severe Accident Technical Guidelines. 5.2.2.13 The Control Parameter Assessment Engineer evaluates the availability of instrumentation used to determine values of the Emergency Operation Procedures/Severe Accident Guideline control parameters. 5.2.2.14 The Security Coordinator provides security plan knowledge and expertise. Coordinates all security related response activities including initial and continuous accountability of personnel when required per EPIP 5.7.10. The Security Coordinator may be assisted by other members of the CNS Security force. EMERGENCY PLAN REVISION 76 PAGE 37 OF 192

5.2.2.15 The Facility Logkeeper maintains an accurate log of important TSC functions and also maintains/updates the display of priority work items. 5.2.2.16 The Administrative Assistant provides administrative support such as faxing, copying, and material needs. 5.2.3 EOF DIRECTOR The EOF Director is in charge of the EOF functions and responsibilities, including ensuring the EOF is capable of supporting the Emergency Director's management of the overall licensee emergency response. The primary functions of the EOF staff are to provide assistance to the Emergency Director, coordination of emergency off-site response activities, and to provide support to the responding off-site support agencies by:

Coordinating radiological and environmental assessment.
Determining and recommending protective actions for the public.
Coordinating emergency response activities with Federal, State, and Local agencies.
Event classification and continual assessment of plant conditions related to classification.
Notification to off-site authorities.

0 The EOF Director is assisted in these functions by individuals assuming the minimum staff ERO positions below and depicted in Figure 5.2-3. 5.2.3.1 The Radiological Control Manager provides radiological information and recommendations to the Emergency Director and/or EOF Director with regard to dose assessment, protective actions, and the use of Potassium Iodide. The Radiological Control Manager is assisted by and directs the activities of the Radiological Assessment Supervisor. Additional duties include interfacing with appropriate State and Local Dose Assessment Groups. 5.2.3.2 The Radiological Assessment Supervisor assists the Radiological Control Manager in determining potential or actual impacts of radiological releases, developing protective action recommendations, and coordinating the activities of the Field Monitoring Teams. This is accomplished by supervising the activities of the Field Team Coordinator and Dose Assessment Coordinator located in the EOF Dose Assessment Room. 5.2.3.3 The Off-Site Communicator is responsible for gathering and disseminating information to appropriate Off-Site Agencies in accordance with EPIP 5. 7 .6. EMERGENCY PLAN REVISION 76 PAGE 38 OF 192

NOTE - The additional Key Functional staff and other positions listed below enhance the operation of the EOF. 5.2.3.4 The Operations/EOP Advisor provides technical assistance and operational information to the Emergency Director and/or EOF Director. 5.2.3.5 The Field Team Coordinator coordinates the movement and sampling activities of the CNS Field Monitoring Teams as directed by the Radiological Assessment Supervisor. 5.2.3.6 The Dose Assessment Coordinator coordinates dose assessment activities as directed by the Radiological Assessment Supervisor. This individual has to be familiar with source term data, release data, meteorological information, and other dose assessment parameters. 5.2.3. 7 The Dose Assessment Clerk performs dose assessment as instructed by the Dose Assessment Coordinator using assessment methods as described in Section 6.3.3. 5.2.3.8 The Field Monitoring Team Vehicle Driver drives the field monitoring team vehicle. 5.2.3.9 The Field Monitoring Teams are composed of at least one individual selected from a pool of personnel knowledgeable and experienced in radiation protection as defined by ANSI Standard 18.1, and trained in sampling techniques and analysis in accordance with the Emergency Preparedness Training Program. They are familiar with the equipment and methods to be used to perform plume-tracking and media sampling due to previous experience in radiological protection. Other CNS personnel may act as vehicle drivers or assistants. 5.2.3.10 The Logistics Coordinator is responsible for providing on-going EOF security and accountability, food/lodging/transportation support, and coordinating the capability of 24 hour continuous operations staffing. 5.2.3.11 The Emergency Preparedness Coordinator assists with activation of the Emergency Response Facilities and ensures that ERO personnel are performing their duties as defined by the Emergency Plan, EPIPs, and Positional Instructional Manuals. 5.2.3.12 The Clerical Coordinator ensures that sufficient clerical support exists in the EOF to adequately support EOF personnel. 5.2.3.13 The Facility Logkeeper maintains an accurate log of all important EOF activities, and also maintains and updates the display of EOF priority work items. EMERGENCY PLAN REVISION 76 PAGE 39 OF 192

5.2.4 OPERATIONAL SUPPORT CENTER (OSC) SUPERVISOR The OSC Supervisor is in charge of OSC functions and activities. His/her primary responsibility is to assure work items assigned to the OSC, based on the direction and priorities established by the TSC and assigned by the Maintenance Coordinator, are carried out.

The OSC is located adjacent to the TSC. Functional assignments at the OSC, coordinated from the TSC are:
Operating staff support.
Radiation surveys and decontamination.
Maintenance, repair, and damage control.
Chemistry.
Re-entry, search, and rescue.

The OSC Supervisor is assisted in these functions by individuals assuming the minimum staff ERO positions described below and depicted in Figure 5.2-4.

Electricians (2).
l&C Technicians (2).
Mechanics (2).
Radiation Protection Technicians (6).

The additional Key Functional staff and other positions listed below enhance the operation of the OSC. OSC Leads listed below work together to assign emergency mitigation work activities to available OSC personnel best suited in performance of the assigned task. The OSC Leads work as a multi-disciplinary team to assemble, brief, and dispatch teams. They are also responsible for monitoring the progress of the respective teams, overseeing their safety, and debriefing them upon completion of their assigned tasks.

Chemistry/Radiological Protection Lead.
l&C Lead.
Electrical Lead.
Mechanical Lead.
Utility Lead.
An OSC Clerk provides clerical support such as logkeeping, faxing, copying, and material requisition, to the OSC Staff.
The Technical Communicator provides a flow of technical data from the Technical Communicator in the Control Room to the OSC.

EMERGENCY PLAN REVISION 76 PAGE 40 OF 192

The OSC also contains a pool of trained personnel with expertise from their normal day-to-day activities. The following are examples of these additional personnel from which teams may be assembled:

Welders/Pipefitters/Machinists.
Chemistry Technicians.
Utility/Tool Crib.
Warehouse Personnel.
Operators.
Engineers.

5.3 OFF-SITE EMERGENCY ORGANIZATION The Emergency Plan is designed to be implemented in a step-by-step fashion as site needs dictate. The off-site capabilities activated by this plan will have pre-assigned duties meant to relieve site personnel of off-site related responsibilities as soon as practical. This shifting of responsibilities will take place rapidly and formally as the emergency evolves and will relieve site personnel needed for in-plant activities. NPPD employees located at the General Office or other NPPD facilities may be used to form a technical manpower pool from which technical support may be drawn. These employees may be utilized by virtue of their normal job position, availability, or personal qualifications. 0 5.3.1 JOINT INFORMATION CENTER (JIG) The JIG is a media briefing area and is located adjacent to the EOF at 902 Central Avenue in Auburn, NE. The principal functions of the JIG include:

Coordinating the development and dissemination of information to the *news media.
Conducting media monitoring.
Maintaining rumor control.
Providing NPPD employees with information concerning the emergency.

The staffing of the JIG will be dependent upon the type of emergency situation at CNS. A minimum staffing level, described below and depicted in Figure 5.3-1, will ensure principal functions of the JIG can be accomplished: 5.3.1.1 The JIG Director directs personnel in preparation of position statements, interviews, and dissemination of information to employees, participants, industry organizations, legislative representatives, and members of the Board of Directors. He/she is also responsible for generating news releases and is responsible for ensuring that the information authentication function is being performed. EMERGENCY PLAN REVISION 76 PAGE 41 OF 192

5.3.1.2 The Technical Briefer will assist the JIC Director by receiving and relaying technical information. He/she is also responsible for advising the JIC Director in matters regarding Tech Specs, USAR, EOPs and EPIPs. 5.3.1.3 The Public Information Officer prepares releases for the news media and provides support to the Designated Spokesperson. He/she also coordinates with Public Information Officers from other agencies, responds to inquiries from the public, and assists with other JIC activities as necessary. To enhance the effectiveness of the JIC, the JIC Management Staff is also supported by the following Key Functional and other positions:

Facility Manager.
Media Monitor.
Designated Spokesperson.
Rumor Control Coordinator.
Employee Information Coordinator.
JIC Clerical Coordinator.
JIC Logkeeper.
Rumor Control Staff (NPPD Centralized Customer Care Center).

0 5.3.2 PUBLIC INFORMATION SUPPORT Emergency public information will be coordinated and released through the Joint Information Center (JIC). Public information releases to the news media will be channeled through the JIC. Accurate and timely information on emergency conditions will be transmitted to JIC personnel. Coordinated news conferences will be conducted by Public Information Officers representing NPPD, as well as Federal, State, and Local Agencies. Provisions are made for a question and answer exchange. The NPPD Designated Spokesperson located at the JIC is responsible for ensuring that information pertaining to events at CNS is properly transmitted to the news media. The Designated Spokesperson or JIC Staff will be in contact with personnel in the EOF and will organize and distribute the technical information for use in media briefings and news releases. EMERGENCY PLAN REVISION 76 PAGE 42 OF 192

5.3.3 CONTRACT SUPPORT In addition to General Office support, the CNS Emergency Response Organization may draw on outside support. Letters of Agreement with organizations which may provide assistance to NPPD are listed in Appendix D. A brief description of this contract support is provided below. 5.3.3.1 MANPOWER AND EQUIPMENT AUGMENTATION The Institute of Nuclear Power Operations, as an organization serving the nuclear industry, has organized a response plan for nuclear power plant emergencies. Manpower and equipment may be requested from institute members to augment on-site capabilities. 5.3.3.2 TECHNICAL SUPPORT The General Electric Company has organized a Boiling Water Reactor Emergency Support Program. This program provides for an Emergency Response Team composed of personnel with appropriate technical disciplines, which will report to NPPD upon request. A Technical Support Team is also established at General Electric Nuclear Headquarters in Wilmington, North Carolina. Communications between the CNS Emergency Response Organization and General Electric will enhance technical assistance to the station. 0 In the event CNS needs additional technical support, NPPD has made arrangements with the following organizations:

POWER Engineers (formerly Bums & Roe).
CB&I - Stone & Webster.

5.3.3.3 RADIOLOGICAL MONITORING AND ANALYSIS SUPPORT Arrangements have been made with Omaha Public Power District's Fort Calhoun Nuclear Station to provide monitoring equipment and personnel trained to use this equipment, contingent on the manpower, equipment, and supplies that are available at the time of the request for assistance. Additionally, Fort Calhoun Station is a decommissioning plant and staffing reductions may limit the total number of personnel available for assistance. The Radiochemistry Laboratory at the Fort Calhoun Station is able to perform backup radioisotopic analyses of monitoring samples. If the emergency is such that more monitoring equipment and personnel are needed, resources such as the Institute of Nuclear Power Operations may be requested. Emergency service is available from the current dosimetry vendor which includes extra dosimetry, instrumentation, and technical assistance. EMERGENCY PLAN REVISION 76 PAGE 43 OF 192

5.4 PARTICIPATJNG FEDERAL, STATE AND LOCAL AGENCIES ~ Figure 5.4-1 depicts the interrelationships among some of the various State and

'       Federal organizations which may respond to an emergency at CNS.

Off-site monitoring and assessment activities will be coordinated at the EOF. The General Office support groups, as well as State, Local, and Federal Agencies will coordinate their efforts through the EOF (Figures 5.4-1 and 5.4-2). The NRG on-site effort may be coordinated through the TSC or EOF, whichever is appropriate. The affected states may send liaison representatives to the EOF to aid in the coordination effort. 5.4.1 THE STATE OF NEBRASKA In the State of Nebraska, the Nebraska Emergency Management Agency, under the Nebraska Adjutant General, is the lead planning agency for developing radiological emergency plans for fixed nuclear facilities. On receipt of information indicating the need for State and Local Government response, a disaster emergency condition will be declared by the Governor and the State Emergency Operations Center will be activated. State agencies having responsibilities under the Nebraska State Radiological Emergency Response Plan for nuclear power plant incidents will be notified and kept informed of the progress of the emergency as discussed in that Plan. A Governor's Authorized Representative (GAR) will be designated by the Governor. The GAR will coordinate activities of state agencies responding to the emergency. The GAR will also be a point of contact for decisions involving implementation of protective actions as recommended by the Emergency Director. The Nebraska State Emergency Operations Center will be the principal point of contact with the Emergency Operations Centers of adjacent states. As conditions warrant, the state EOG command and control functions may be carried out from other designated facilities. A Nebraska Health & Human Services Regulation and Licensure (HHSRL) representative will be at the Field Command Post or the EOF. Acting in coordination with NPPD Management and other agencies, he/she is responsible to perform initial state assessment of the health hazard to include development of recommendations for initiation of protective actions. This individual or representative will also coordinate the activities of the state Radiological Field Monitoring Teams and advise the Governor's Authorized Representative and local governments as to health hazards of the incident. Radiological monitoring will be conducted by both CNS Field Monitoring Teams and the HHSRL Radiological Field Monitoring Teams. The lead agencies for the countywide emergency planning in Nemaha, Richardson, and Otoe Counties are the respective County Emergency Management Agencies/Directors. The responsibilities of various county groups are described in the appropriate annexes of the individual County Radiological Emergency Response Plans. EMERGENCY PLAN REVISION 76 PAGE 44 OF 192

5.4.2 THE STATE OF MISSOURI The principal state agency for the coordination of emergency response in the State of Missouri is the Missouri State Emergency Management Agency (SEMA). SEMA coordinates actions, operations, and resources involving response required to support decisions affecting the emergency. The Missouri Department of Health and Senior Services, through the Bureau of Environmental Epidemiology, is responsible for aU decisions affecting protective responses, dose, dose commitment during the emergency, and recovery in the emergency area. In the event of an emergency, communications between CNS and the Division of Health is maintained in order to confirm measurements and estimates of possible off-site consequences and to keep the State EOG informed of the status of the emergency. Emergency response and support operations will be initiated through decisions made jointly by the Director, SEMA, and the Director, Division of Health or their duly appointed representatives, or on request of affected governmental officials. In the event of an emergency, which may present an off-site hazard to the public, the State Emergency Operations Center at Jefferson City will be activated in accordance with the State Emergency Operations Plan. A representative of the state may be dispatched to the EOF. He will have direct communications with the Forward Command Post to provide accurate and 0 timely information to State and Local Response Forces. Atchison County authorities are notified through the Atchison County 911 Center. The 911 Center is notified by NPPD or by SEMA. The response from these authorities is more fully detailed in the Atchison County Nuclear Emergency Response Plan. 5.4.3 THE STATES OF KANSAS/IOWA The States of Kansas and Iowa may also play an active role in responding to an emergency at CNS. While neither state is within the 10 mile EPZ, they are located within the 50 mile Ingestion Exposure Pathway. As such, NPPD will maintain liaison with the appropriate officials of these states and provide information and recommendations as the situation dictates. More detailed information can be found in the Emergency Response Plan of each respective state. EMERGENCY PLAN REVISION 76 PAGE 45 OF 192 I

5.4.4 NUCLEAR REGULATORY COMMISSION The NRG regulates nuclear activities to protect the health and safety of the public and to preserve environmental quality and has developed an Incident Response Plan to ensure that its statutory responsibilities are fulfilled. The responsibilities assigned by the NRC plan are exercised through a set of implementing procedures that delineate the manner in which each function will be performed, the criteria to be used in making each decision, and the information needed for both. When NPPD notifies the NRG of an emergency, the initial NRC response is to ascertain the status of the station and monitor emergency response activities to assure that the public and the environment are fully protected. The NRC will measure off-site radiological effects and develop projections of on-site and off-site effects for the use of other federal, state, and local agencies. The NRG may offer specific advice to NPPD to help solve or limit the consequences of the problem. The NRG is prepared to amend or change CNS Technical Specifications or to issue formal orders if NPPD should fail to take whatever actions the NRG deems necessary to Rrotect the public. The Chairman of the Commission is the senior NRG authority for all aspects of emergency response and will become the "Director" of all NRG activities and personnel. Normally, the Chairman will delegate responsibilities to a "Deputy Director" 0 upon activation of the Operations Center. The Deputy Director will carry out the delegated responsibilities unless the Chairman specifically directs otherwise. Together, the Director and Deputy Director assure that preplanned actions commence and identify other necessary actions unique to the particular incident. Headquarters and region teams will carry out these actions. The Director may appoint an NRG "Director of Site Operations" as soon as a qualified official arrives at the site, assesses the situation, and reports back to the Director. The Director may also delegate one or more of the following authorities to the Director of Site Operations:

Authority to recommend actions to the licensee.
Authority to direct the licensee to take specific actions.
Authority to recommend actions off site, including protective measures for the public.

Other officials and organizations will be immediately informed of the appointment and delegated authority. The Director of Site Operations will assume supervision of all NRC personnel at the site, will represent the NRG in interactions with other agencies, and will decide what response actions must be taken, consistent with the delegated authority. EMERGENCY PLAN REVISION 76 PAGE 46 OF 192

Figure 5.2-1 CNS Normal Operating Organization - Control Room POSITION : Number Shift Manager (1) Control Room Superviso r (1) Reactor Operator (3) Station Operator (3) Chem/RP Technician ( 1) Shift Communi cator ( 1) Dose Assessor ( 1) Utility/Fire Brigade (2) Shift Technical Engineer (1)+ Control Room Logkeepe r (1 )* Technical Communic ator (1 )* Shift Dose Control Control Shift

                                                         . Utility/Fire

+ The Shift Technical Engineer is not required to be on-shift during cold shutdown conditions.

These personnel are not on-shift.

Note - Figure 5.2-1 is based on the staffing analysis required by 10CFR50 Appendix E, IV.A.9, by Decembe r 24, 2012. NRC NSIR/DPR-ISG-01 , "Interim Staff Guidance" and NEI 10-05, "Assessm ent of On-Shift Emergenc y Response Organization Staffing and Capabilitie s" , was used to conduct and document this analysis . Figure 5.2-1 new staffing numbers become effective on January 19, 2013 . EMERGENC Y PLAN R EVI SION 76 PAGE 47 OF 192

Figure 5.2-2 CNS Emergency Response Organization - Technical Support Center (TSC) r-'\I POSITION: TSC Director1 Operations Coordinator1 1 Chemistry/R adiological Protection Coordinator Maintenance Coordinator1 Engineering Coordinator1 Operations/E OP Advisor2 Security Coordinator2 Engineering Team Leader2 Control Parameter Assessment Engineer2 Function Status Assessment Engineer2 Engineering Staff3 ENS Communica tor2 Administrativ e Assistant TSC Logkeeper TSC Directer TSC ,.. Logkeqxr I I I I I

           ~             Ckm'RP         ENS              Maimemnre        SroJrity        ~                Mni.nilfraive ComfillPler  Coonfimter   Coonnmicater         Coo1finaer     Coo1finater      ~er                ~mt I                                                                                I
            ~                                               osc                           ~

Advi~ Teamleada-I I I I

                                                                         ~               Fuoction Status  Cairol P.iamettr Staff       ~Fnginecr        ~tFliginetr 1 Minimum Staff

'0 2 Key Functional Staff 3 Engineering disciplines will be Reactor, Civil, Mechanical, and Electrical EMERGENCY PLAN REVISION 76 PAGE 48 OF 192

Figure 5.2-3 CNS Emergency Response Organization - Emergency Operations Facility (EOF) POSITION: Emergency Director1 EOF Director1 Radiological Control Manager1 Radiological Assessment Supervisor1 Off-Site Communicator1 Operations/EOP Advisor2 Emergency Preparedness Coordinator2 Dose Assessment Coordinator Field Team Coordinator Dose Assessment Clerk Logistics Coordinator EOF RP Pool EOF Logkeeper Clerical Coordinator Field Monitoring Team Vehicle Driver Emergency Director EOF Dlrector

                                            ,___,,_---, Emergency Preparedness EOF Logkeeper Coordinator Ops/EOP      Off-Site             Radiological                Logistics  Clerical AdVISOf   Communicator             Control                 Coordinator Coordinator Manager Radiological Assessment Supervisor Dose Assessment                  Field Team Coordinator                 Coordinator Dose Assessment                OSC RP Pool Clerk                   Field Monrtoring Teams 1 & 2 Field Monitoring Team Vehlde Drivers 1

Minimum Staff 2 Key Functional Staff EMERGENCY PLAN REVISION 76 PAGE 49 OF 192

Figure 5.2-4 CNS Emergency Response Organization - Operations Support Center (OSC) POSITIONS: OSC Supervisor2 Technical Communicator Chem/RP Lead 2 RP Technicians 1 (6 minimum) Electrical Lead Electricians 1 (2 minimum) Mechanical Lead Mechanics 1 (2 minimum) l&C Lead l&C Technicians 1 (2 minimum) Utility Lead Utility personnel OSC Clerk Warehouse person 0 Maintenance Coordinator IOSC Supervisor I Technical OSC Clerk - - Communicato r I I I I I I Chem/RP I&C Electrical Utility Mechanical Warehouse Lead Lead Lead Lead Lead I I I I I RP Technicians I&C Electricians Utility Mechanical Tecbnician.5 1 Minimum Staff 2 Key Functional Staff EMERGENC Y PLAN REVISION 76 PAGE 50 OF 192 I

Figure 5.3-1 CNS Emergency Response Organization - Joint Information Center (JIC) POSITION: JIC Oirector1 Technical Briefer1 Public Information Officer1 Facility Manager2 Designated Spokesperson 2 Rumor Control Coordinator2 Employee Information Coordinator2 JIC Logkeeper JIC Clerical Coordinator Media Monitor Rumor Control Staff (NPPD CCCC) JIC Director JIC JIC Clerical 1-----'1'--------1 Logkeeper Coordinator I Designated Public Information Facility Media Manager Monitor Spokesperson Officer I I Technical Rumor Control Employee Information Briefer Coordinator Coordinator I Rumor Control Staff NPPD CCCC 1 Minimum Staff 2 Key Functional Staff EMERGENCY PLAN REVISION 76 PAGE 51 OF 192

Figure 5.4-1 Interrelationships of Emergency Response Organizations NRC (3) Control Room NRC (2) Operations Technical Support

                   "'l'---------t                    Center                               ----                                Support Center ON-SITE OFF-SITE                              ' ---         -                -- --- --7                            --           -          -                -      .,

NRC (2) ~ n::.t ,21:: *; \' !

                                                                                                                     ~ i 'I

IL,',., * ,; '4 I * ;

J; ! ::le~,..-:-: l: f I ,. :;_ :

                                                   , -;i: ,I; .'                         .
                                                                                                                                    -: ,.~1                     l Federal                                                           . 1;.l11  \l:13) .. , 1.) ;$   ,;

J Support (1) :..--:q'f: ; ) - ~ - - - ~ State(s} EOC County(s) EOC Support NPPD NPPD/Agency Agency Support Notes: (1) See Figure 5.4-2 for detailed information on Federal Support. (2) NRC Support in TSC and EOF. (3) NRC Resident Inspector located in the Control Room. EMERGENCY PLAN REVISION 76 PAGE 52 OF 192

Figure 5.4-2 Federal Response Management Diagram Cooper Nuclear Station IGovernor or Designated Representative I NRC/FEMA Joint I Coordination I I I NRG Coordinates FEMA Coordinates Technical Aspects of Non-Technical Aspects Federal Response of Federal Response I I I NRC DOE Coordinates NCS Federal Offsite >- Radiological Monitoring I DOD I Utility I HNRG/Utility I >-

                                       ~       EPA       I                  DOC H DHHS            I H USDA            I      -

USDA

                                       ~      DOC        I DHHS H      DOE        I      -

y DOD I FEMA EPA DOT Source: Federal Register 45FR84911 EMERGENCY PLAN REVISION 76 PAGE 53 OF 192

6. EMERGENCY MEASURES Nebraska Public Power District (NPPD) emergency measures will be conducted in accordance with the particular emergency classification at Cooper Nuclear Station (CNS).

This section of the plan (1) discusses emergency alarms and evacuation, (2) identifies segments of the station emergency organization that will be activated at each class of emergency, (3) details methods and procedures for assessment actions, (4) specifies actions to correct or minimize the emergency situation, (5) describes protective actions to prevent or minimize radiological exposure, and (6) discusses aid to affected personnel. 6.1 SITE EMERGENCY ALARMS When an emergency condition exists that could affect the safety of station personnel, the appropriate alann will be manually activated and an announcement made from Control Room. If the condition involves a fire, the fire alann will be activated and designated personnel will respond. If the condition results in an emergency declaration, the emergency alarm will be activated and an announcement made. The condition of the emergency will dictate what directions will be given during the announcement. Site Security personnel may assist in notification of personnel on NPPD property. 6.2 NOTIFICATION AND ACTIVATION OF EMERGENCY RESPONSE ORGANIZATIONS The four classes of emergencies defined in Section 4 require a varying degree and scope of emergency response. The appropriate parts of the emergency response organization activated in each emergency classification are presented in Section 5. The transition from the nonnal operating organization to the emergency response organization involves the following steps:

Notify the emergency response organization members who are off-site, or are on-site but may not be aware of the emergency, that their assistance is required.
Fill emergency response positions on an interim basis with personnel who are immediately available at the time of the emergency.
Fill positions in the emergency organization with ERO members as they arrive at the various Emergency Response Facilities.

6.2.1 ON-SITE PLANT PERSONNEL Plant personnel on-site are notified by an emergency alarm and announcement as described in Section 6.1 and EPIP 5.7.2. 6.2.2 OFF-SITE PLANT PERSONNEL Plant personnel (ERO) off-site are notified in accordance with EPIP 5.7.2. This is normally accomplished via the CNS Automated Notification System. A listing u of telephone numbers for notification of ERO members is maintained by the Emergency Preparedness Department. EMERGENCY PLAN REVISION 76 PAGE 54 OF 192 j

6.2.3 JOINT INFORMATION CENTER (JIC) EPIPs 5. 7.2, Emergency Director EPIP, and 5. 7.6, Notification, define how General Office personnel are notified. Depending upon the situation, the JIC may be activated. Personnel notification schemes and procedures for activating this facility are contained in EPIP 5.7.23. 6.2.4 OFF-SITE AUTHORITIES AND SUPPORT AGENCIES When an emergency classification is declared, CNS will initiate pre-determined notifications as defined in EPIP 5.7.6. Initial notifications to responsible State and Local Governmental Agencies will be completed within 15 minutes of the declaration of an emergency. The contents of initial and follow-up notifications are set forth in EPIP 5.7.6, and contain information about the class of emergency, release information, potentially affected population areas, and protective action recommendations. Follow-up communications with off-site authorities will consist of periodic messages containing additional information as described in EPIP 5.7.6. Notifications to responsible State and Local Governmental Authorities are normally accomplished via the CNS State Notification Telephone System. Off-site authorities, as well as technical support groups likely to be consulted in an emergency, are listed in the Emergency Telephone Directory. 6.2.5 NUCLEAR REGULATORY COMMISSION (NRC) Notification of the NRC Operations Center for all emergency classification levels is normally accomplished via the Federal Telecommunication System Emergency Notification System (FfS-ENS) in accordance with EPIPs 5.7.6 and 5.7ENS. EMERGENCY PLAN REVISION 76 PAGE 55 OF 192

6.3 ASSESSMENT ACTIONS The assessment of station conditions, radiation levels, and off-site consequences is initially conducted by the Control Room. Radiological dose assessment can also be performed in the Control Room, as necessary. The Shift Manager, in the role of Emergency Director, activates the emergency response organization described in Section 5 per EPIP 5.7.2. Assessment actions described in Table 6.3-1 will continue throughout the emergency. These assessments may result in reclassification, which could alter emergency response actions. CNS has systems for monitoring radioactive materials released to the environment, and is equipped with process and system monitors capable of assessing radiological conditions and initiating appropriate alarms or actuating control equipment for containment of radioactive materials if pre-established limit~ are reached. These systems will monitor radioactive releases during accident conditions. 6.3.1 POST-ACCIDENT SAMPLING SYSTEM Samples of both reactor water coolant and drywell atmosphere can be drawn using Post-Accident Sampling System. This system allows personnel to safely take samples and conduct analyses, while keeping radiation dose to personnel within specified limits. Samples collected and analyzed will provide information, which may indicate reactor conditions such as cladding failure, effects from high fuel temperature, or fuel melting. () 6.3.2 METEOROLOGICAL DATA The site has meteorological instrumentation, which indicates and records wind speed, wind direction, and temperature differentials on a continuous basis. Detailed information on this system can be found in Section 7.5.2. A continuous readout of this information is available on the Plant Management Information System (PMIS). In the event that meteorological information from this primary source is unavailable, meteorological information is available from the National Weather Service. 6.3.3 DOSE ASSESSMENT CNS has the capability of performing dose projections during a radiological emergency using two separate techniques:

CNS-DOSE: CNS-DOSE, a computerized class 'A' model, is the primary method of performing rapid dose projections (predictions) in order to develop protective action recommendations during the early accident phase within the plume exposure EPZ. The program is operated on plant computers and can make use of current meteorological and radiological effluent monitor readings as well as manually entered data.

EMERGENCY PLAN REVISION 76 PAGE 56 OF 192

HAND CALCULATION: A manual calculation, derived from the methodology utilized by CNS-DOSE, is the backup method for performing dose projections In order to develop protective action recommendations during the early accident phase within the plume exposure EPZ. Input data is taken from the same resources as the computerized method. Both centerline and off-centerline calculations can be performed.

Once the EOF is operational, dose projection and assessment responsibilities are transferred from the Control Room to this facility. The TSC also has the capability to perform dose assessments. Off-site concentrations of radionuclides and radiation dose rates are determined by NPPD and State Field Monitoring Teams. Once the EOF is operational, Field Monitoring Teams are deployed with portable radiological instrumentation (air samplers and radiation survey meters) and communications equipment in vehicles designated for this purpose. The field instrumentation used for airborne activity monitoring has the capability to detect radioiodine concentrations as low as 1.0x10-7 µCi/cc (microcuries per cubic centimeter). Field information is used to validate dose projections and to assist in determining the adequacy of protective actions. 6.4 CORRECTIVE ACTIONS Instrumentation and Control Systems monitor, provide indications and alarms, record, and automatically control systems necessary for the safe operation of the station. 0 Control and display of information from these systems is centralized in the Control Room. Displays are also available in the TSC and EOF. This instrumentation is a source of information used to determine emergency classification as shown in Tables 4.1-1 through 4.1-4, and EPIP 5.7.1 (Emergency Classification), and may provide entry conditions for Abnormal Operating Procedures (AOPs), Emergency Operating Procedures (EOPs) or Severe Accident Guidelines (SAGs). AOPs, EOPs, and SAGs contain steps for preventive and/or corrective actions to avoid or mitigate consequences of an emergency. During a declared emergency, corrective actions are performed by the ERO under the direction of the Emergency Director. These corrective actions are designed to (1) terminate the accident, (2) mitigate or eliminate potential hazards to the public and station personnel, (3) restore the plant to a safe and stable condition, and (4) de-escalate the emergency classification. The potential nature of some emergencies may warrant the utilization of off-site individuals, organizations, and agencies. As a result, local support service arrangements have been made with off-site groups to provide on-site aid in the event of an emergency situation, including those resulting from hostile actions. Corrective actions may also involve response by the following: EMERGENCY PLAN REVISION 76 PAGE 57 OF 192

FIRE BRIGADE The CNS Fire Brigade will respond to station fire calls. The Fire Brigade is composed of the Fire Brigade Leader, two Station Operators, and two other individuals qualified as Fire Brigade members, in accordance with station procedures. If off-site firefighting assistance is required, including that caused by hostile action, fire response including fire apparatus and firefighters will normally be requested by the CNS Control Room to the Nemaha County 911 Dispatch Center and implemented using the National Incident Management System (NIMS).

Mutual aid may also be requested via the Nebraska Emergency Management Agency. Off-Site Fire Department(s) will be escorted to the fire scene by Security personnel upon arrival.

REPAIR AND DAMAGE CONTROL TEAM For minor emergencies, station personnel will handle cleanup, repair, and damage control. For more severe emergencies, the support of additional NPPD personnel or specialized outside contractors may be required to assist in damage control, cleanup, and repair operations.

6.5 PROTECTIVE ACTIONS On-site actions to protect station personnel and visitors during a declared emergency are the responsibility of the Emergency Director. Measures for the protection of the general public are detailed in the State Emergency Response Plans. 0 Protective actions for on-site personnel will be taken whenever a radiological emergency has occurred, or may occur, which might result in concentrations of airborne activity or radiation levels in excess of pre-determined limits. Protective actions will also be taken for on-site personnel in other emergency situations such as fires, floods, tornadoes, or security related events where personnel safety is threatened. A range of protective actions to protect on-site personnel during hostile action is provided to ensure the continued ability to safely shut down the reactor and perform the functions of the emergency plan. An alternative facility, with communication capabilities for contacting the Control Room, plant security, and the EOF is available to serve as a staging area for Augmented Emergency Response Staff if the site is not accessible or the site is under threat _of or experiencing hostile action. Activation of Emergency Response Facilities, assembly and accountability activities, and evacuation of site personnel may be delayed if it is determined by the Emergency Director that personnel safety would be threatened. In this situation, on-site personnel will be notified of these events by the station alarms, telephone calls, or public address system announcements, as applicable. Personnel will be notified of appropriate protective actions to be taken as soon as assessment actions permit a proper evaluation of conditions. EMERGENCY PLAN REVISION 76 PAGE 58 OF 192

Following the instructions and using the procedures referenced in EPIP 5.7.20, the Radiological Control Manager will determine if the projected downwind doses indicate a need to implement any type of protective actions. If the results of the analysis indicate a need to implement protective actions, he will inform the Emergency Director of his findings and together they will decide if protective action recommendations are warranted. The Emergency Director will provide protective action recommendations to off-site authorities. Protective actions for off-site areas are implemented by State and Local Government Emergency Response Organizations. These actions may include evacuation or in-house shelter. Factors such as release duration, mobilization time, or adverse weather will be important considerations affecting protective actions. The action which affords the lower radiation dose is preferred. Approximate initiation times for protective actions are shown in Table 6.4-2. Within the Plume Exposure Pathway, an Alert and Notification System (ANS) has been installed. Residents of this EPZ have been instructed to tune to their local Emergency Alert System radio station for further instructions when the ANS is activated. This system was established to meet the prompt notification requirements established by the NRG and was designed for response to any disaster where prompt notification of the public is desirable. The design basis and rationales for the ANS is in the ANS Design Report. Details pertaining to physical and administrative controls of this system are also found in this document. The ANS includes fixed sirens and digitally-activated National Oceanic and (J Atmospheric Administration/Emergency Alert System (NOAA/EAS) Radio Receivers. These radio receivers are made available to residences located within the Plume Exposure Pathway, but outside the hearing range of the fixed sirens. The radios are pre-tuned to an EAS station and are automatically activated when the EAS is activated. Special use or remote area notification is discussed in the CNS ANS, Design Report. State and Local Plans have provisions for notifying the transient population within the Plume Exposure Pathway. EPIP 5.7.27 describes how the system will be activated in the event of an emergency. Local and/or state governments are responsible for implementation of notification/Warning actions. Normally, public warning information will be disseminated as directed by the Governor or his Authorized Representative. However, the Notification System provides for local government decision and initiation of notification/warning actions, especially in the event of a major nuclear power plant incident. Local governments may make decisions based on the recommendations of NPPD Management or state representatives. As indicated in EPIP 5.7.27, pre-arranged messages are used when instructing the general public on what actions should be taken. Approximately 15 minutes will be required to notify the public from the time the decision has been made to activate the system to the time required to broadcast a message. EPIP 5.7.27.1 describes how a malfunctioning NOAA/EAS radio will be repaired or replaced. EMERGENCY PLAN REVISION 76 PAGE 59 OF 192

Back-up ANS methods have been established and are implemented by the Local Emergency Management Agencies. The back-up ANS for Atchison County, Missouri is Route Alerting. Route Alerting maps can be found in Atchison County Transportation and Fire Department procedures. Atchison County also uses social media to provide emergency alerts to the public, as a back-up method. Instructions for Route Alerting and the use of social media can be found in the Atchison County procedures. The 10-mile EPZ population in Missouri is approximately 2,215 people based on the current Cooper Nuclear Station Evacuation Time Estimate. Based on the guidance contained in the above procedures, there is reasonable assurance that back-up ANS can be completed within the target time of 45 minutes. The back-up ANS for Nemaha and Richardson Counties in Nebraska is the cloud-based mass notification service "Southeast Nebraska Emergency Notification System (SEN ENS)". This system can send emergency messages to geo-coded (by address) telephones throughout the EPZ. Otoe County Emergency Management Agency is the sponsor of this system. The primary activation points for this system are located at the Nemaha County Emergency Management office, the Richardson County Emergency Management office, and the Otoe County Emergency Management office. Multiple activation points offer the advantage of back-up activation capability should one activation point lose activation capability. The 10-mile EPZ population in Nebraska is approximately 1,800 people based on the current Cooper Nuclear Station Evacuation Time Estimate. Thus, including the potential addition of households that have land-lines and cellular communication voice or SMS text devices, plus business and institutional addresses correspond to a maximum expected number of contacts to be made, there is reasonable assurance that back-up ANS can be completed on an area-wide bases within the target time of 45 minutes. 0 6.5.1 RESCUE OPERATIONS The search and rescue function is handled by trained emergency response personnel. If station personnel are unaccounted for in the initial or subsequent personnel accountability, an emergency team will be assigned to locate and, if necessary, rescue them, observing the guidelines set forth in EPIP 5.7.15. 6.5.2 ON-SITE PROTECTIVE EQUIPMENT AND SUPPLIES Protective equipment is available on-site to minimize radiological dose and contamination, as well as firefighting hazards. The types of equipment include full-face particulate respirators, self-contained breathing apparatus, protective clothing, and air-fed respirators. This equipment is located in the normal station storage areas, and in or near the Control Room and the Emergency Response Facilities. An inventory of this equipment is contained in EPIP 5.7.21. EMERGENCY PLAN REVISION 76 PAGE 60 OF 192

6.5.3 PERSONNEL ASSEMBLY AND ACCOUNTABILITY When the emergency alarm is sounded, site visitors, contractors, and non-ERO on-site personnel will proceed to their Designated Assembly Area. Emergency response personnel will report to their Emergency Response Facility or designated assembly area. The results of personnel accountability will be compiled and reported to the Emergency Director. In the event an individual cannot be located, search teams will be dispatched. Initial accountability will be completed within 30 minutes and continuous accountability will be maintained throughout the course of the emergency per EPIP 5.7.10. 6.5.4 DISMISSAL AND EVACUATION When the emergency alarm is sounded, all personnel will proceed to their designated assembly area by the most direct route unless otherwise instructed to avoid specified areas. The classification and magnitude of the emergency will dictate which Emergency Response Facilities shall be activated, the areas of the site to be avoided, and the off-site support required. Upon assembly and accountability of all personnel, it may be appropriate to dismiss specific personnel to go home, or necessary, to direct personnel to the alternate assembly point located in the Nemaha County Maintenance Facility. All SITE AREA and GENERAL EMERGENCIES require evacuation of all non-ERO personnel. EPIP 5. 7.11 provides the specific procedures to be followed in the event site dismissal or evacuation is required. Once the decision has been made to evacuate, non-ERO employees, contractors, and visitors can be evacuated and relocated to a remote assembly area within approximately one hour. Personnel will not return to the station or deactivate Emergency Response Facilities until directed by the Emergency Director or until the "ALL CLEAR" signal is sounded by the Control Room. 6.5.5 CONTAMINATION AND DOSE CONTROL MEASURES 6.5.5.1 ON-SITE Measures will be taken to prevent ingestion of radioactive materials deposited within the Site Boundary. Affected areas will be isolated. Details of contamination control measures for on-site areas are contained in Station Operation procedures. The monitoring of the work environment within radiological control areas, including specific instructions, precautions, and limitations to personnel working within these areas is supervised by Radiological Protection personnel. Food for emergency response personnel will be provided from off-site sources or from on-site supplies stored in a contamination resistant location. EMERGENCY PLAN REVISION 76 PAGE 61 OF 192

Exposure to airborne radioactivity will be controlled in accordance with appropriate ALARA principles. Periodic air samples will be taken to assure that radioiodine and airborne contamination levels are known. Radio protective tablets (Potassium Iodide) are available for voluntary use by NPPD personnel. NPPD may also provide radioprotective tablets to non-NPPD emergency response organizations for distribution to their emergency workers (i.e., the requesting organization has inadequate supplies). Administration of these tablets to non-NPPD personnel will be the responsibility of these non-NPPD organizations. Any NPPD distribution of radioprotective tablets will be made at the direction of the Emergency Director. EPI P 5. 7 .14 provides further information on the use and distribution of radioprotective (Potassium Iodide) tablets. During an emergency, equipment and tools will be unconditionally released for use outside the area only if their radiation levels are less than 1 mrem/hr above background (fixed contamination) and 220 dpm/100cm2 alpha activity or 2200 dpm/100cm2 beta-gamma activity above background (removable contamination). 6.5.5.2 OFF-SITE For areas beyond the site boundary, Nebraska and Missouri 0 Radiological Monitoring Teams, in coordination with CNS Monitoring Teams, will identify contamination and radiation levels. For areas where public access normally occurs, criteria for off-site areas will be applied. Criteria and measures for contamination control in off-site areas are detailed in the Nebraska and Missouri Emergency Plans. 6.5.6 SECURITY AND ACCESS CONTROL The CNS Security Plan is approved by the NRC and restricts access to the site. Security personnel control access to the Protected Area and during declared emergencies, control access to areas of the Owner Controlled Area. A roadblock will be established on the Plant Access road. Personnel attempting to access the site will be informed of the situation and, rf cleared, will be directed to the proper location. For Security related emergencies, Local Law Enforcement Agency assistance may be requested. 6.6 AID TO AFFECTED PERSONNEL 6.6.1 EMERGENCY PERSONNEL DOSE CRITERIA Dose records for station personnel are maintained by the Radiological Protection Group and are accessible at the TSC. This information will be utilized in determining emergency team assignments. Criteria used for limiting EMERGENCY PLAN REVISION 76 PAGE 62 OF 192

dose to emergency workers are based on recommendations of the U.S. EPA and are shown in Table 6.4-1. n Emergency workers will wear dosimetry as required by Radiological Protection personnel. Emergency worker dosimetry will be provided on a 24-hour basis by Radiological Protection personnel. Every effort will be made to minimize emergency worker dose through the use of protective equipment and supplies, and by minimizing exposure time. The Chemistry/Radiological Protection Coordinator and the Maintenance Coordinator, with assistance of the OSC Supervisor, are responsible for making emergency team assignments. Only the Emergency Director may authorize emergency workers to receive dose in excess of 10CFR20 occupational limits. Personnel conducting corrective or protective actions or life-saving actions who may receive dose in excess of occupational limits should be selected from those who volunteer. Radiological Protection personnel are also responsible for providing self-reading and permanent dosimetry devices to emergency personnel assembled at the OSC and for assuring accountability of each worker's dose. Emergency radiation exposure is controlled in accordance with EPIP 5.7.12. 6.6.2 DECONTAMINATION AND FIRST AID Provisions have been made to assist personnel who are injured, contaminated, or who may have received high radiation doses. Station personnel are trained in first aid and portable first aid kits are available at strategic locations throughout the station. In addition, first aid lockers and decontamination 0 facilities are provided within the station. In the event the above are not available during an emergency, the CNS Communications Building will be used as a first aid station and personnel decontamination center. Detailed information on personnel monitoring and decontamination, including radiological criteria is contained in Station Radiological Protection Procedures. Personnel found to be contaminated will undergo decontamination under the direction of Radiological Protection personnel. 6.6.3 MEDICAL TRANSPORTATION A station ambulance is available for the transportation of injured personnel when station EMTs are available. This vehicle is reserved for emergency use to assure ready availability in time of need. During times when the station ambulance is unavailable due to the performance of preventative maintenance, malfunction, or other circumstances, backup medical transportation services shall be contacted and services requested when necessary. The CNS ambulance is equipped with all required equipment and supplies required by the State of Nebraska for a licensed ambulance. The ambulance also has communications equipment for communicating with the Plant and Nemaha County Hospital. Ambulance attendants (Nebraska-certified EMTs or Paramedics) are trained to handle contamination cases. In addition, NPPD Radiological Protection personnel will accompany contaminated patients to the hospital. If off-site emergency medical response is required, including that caused by hostile action, emergency medical services including ambulances, EMERGENCY PLAN REVISION 76 PAGE 63 OF 192

and emergency medical technicians as requested by the CNS Control Room or TSC to the Nemaha County 911 Dispatch Center and implemented using the National Incident Management System (NIMS) has been arranged. Mutual aid may also be requested via the Nebraska Emergency Management Agency. Off-site medical transport services will be escorted to the scene by Security personnel upon arrival. 6.6.4 MEDICAL TREATMENT FACILITIES The Shift Manager or his designee in accordance with EPIP 5.7.24 will notify the appropriate hospitals if injured personnel are to be transported from the site. Arrangements have been made with the hospitals listed below for care of injured personnel, including cases involving radiological contamination and radiation over-exposure. Selection of the hospital and medical assistance will be based on:

Obtaining the most rapid access to the necessary medical services and facilities.
Capabilities of the specific hospital to provide the required services.
Accessibility due to weather and road conditions.
Number of injured personnel to be transported.
Preference of the injured personnel, if the type and severity of the injury permit.

Except as otherwise dictated by the above conditions, hospitals and medical assistance will be utilized in the following order of priority, based on proximity to Cooper Nuclear Station:

Nemaha County Hospital, 2022 13th Street, Auburn, Nebraska.
University Nebraska Medical Center, Center for Clinical Excellence, East Side, 44th and Dewey, Omaha, Nebraska.

In accordance with Nebraska Statutes (85-805, 806, 807), Nebraska Public Power District supports the University of Nebraska Radiation Health Center and maintains the right to all services of the Radiation Health Center for, but not limited to:

Specialized medical and related services for evaluation, treatment, and management of radiation casualties.
Routine medical, radiation protection, consultation, and associated services.
Educational programs for nuclear safety, with emphasis on preventive medicine and radiological protection.

Patients with radiological injuries beyond the scope of the local medical facilities will be transferred to the Radiation Health Center at the discretion of the local medical staff. EMERGENCY PLAN REVISION 76 PAGE 64 OF 192

C TABLEO1 J ASSESSMENT ACTIONS ACTION DESCRIPTION The radiation level, pressure, temperature, flow, and meteorological data are monitored. Control Room Operators can assess plant status by observing sensor readout. Most sensors

1. Surveillance of Emergency have visual and audible alarms. Data will be provided to the Emergency Director, as Assessment Instrumentation necessary, for his assessment. Control Room Operators will take corrective actions as necessary.

Personnel Assembly and Accountability is the responslbllity of the Security Coordinator and is

2. Personnel Accountability carried out utilizing the security computer. Personnel accountability is maintained by communications with lead personnel at the various Emergency Response Facilities.

Radiological Protection Teams perform these surveys. The radiation levels on the station's fixed area and Ventilation Monitoring Systems will be used to assist in these evaluations.

3. In-Plant Radiological Surveys Contamination surveys of equipment and personnel are conducted with portable equipment from the emergency kits or routine station equipment storage areas.

The surveys are handled by Radiological Protection Teams in same fashion as in-plant

4. Site Boundary Surveys surveys.

Radiological Assessment personnel will use the computer model, effluent monitors,

5. Off-Site Consequence meteorological output, or data supplied by deployed Radiological Protection Teams. Manual Assessment Dose assessment techniques are available in the event computer programs are unavailable.

Samples of various environmental media are collected and analyzed by station Chemistry and Radiological Protection personnel. Results will be evaluated by station personnel with

6. Environmental Monitoring assistance from a contract laboratory if required. State and Federal response personnel may also analyze collected media.

In the case of actual or potential off-site consequences, the State and Local Authorities are immediately notified in accordance with the appropriate CNS EPIPs. Local Authorities use

7. Assessment Reporting pre-determined criteria to initiate various protective actions for the public, as Illustrated in Table 6.4-2.

EMERGENCY PLAN REVISION 76 PAGE 65 OF 192

TABLE 6.4-1 EPA Protective Action Guides (PAGs) for the Early Phase of a Nuclear Incident PROTECTIVE ACTION PAG (PROJECTED DOSE) COMMENTS Evacuation (or, for some Evacuation 1-5 remb situations, sheltering 8 ) should (or sheltering 8 ) normally be initiated at 1 rem. Administration of stable Requires approval of State 25 remc iodine Medical Officials. a Sheltering may be the preferred protective action when it will provide protection equal to or greater than evacuation, based on consideration of factors such as source term characteristics, and temporal or other site-specific conditions. b The sum of the effective dose equivalent resulting from exposure to external sources and the committed effective dose equivalent incurred from all significant inhalation pathways during the early phase. Committed dose equivalents to the thyroid and to the skin may be 5 to 50 times larger, respectively. c Committed dose equivalent to the thyroid from radioiodine. NOTE - From EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Accidents (Table 2-1 ). GUIDANCE ON DOSE LIMITS FOR WORKERS PERFORMING EMERGENCY SERVICES DOSE LIMIT8 (rem) ACTIVITY CONDITION 5 All N/A 10 Protecting valuable property Lower Dose Not Practicable Life Saving or Protection of 25 Lower Dose Not Practicable Large Populations Life Saving or Protection of Only on a Voluntary Basis to

               >25               Large Populations from             Persons Fully Aware of the Extensive Exposure                 Risks Involved a   Sum of external effective dose equivalent and committed effective dose equivalent to non-pregnant adults from exposure and intake during an emergency situation. Workers performing services during emergencies should limit dose to the lens of the eye to 3 times the listed value and doses to any other organ (including skin and body extremities) to 10 times the listed value. These limits apply to all doses from an incident, except those received in unrestricted areas as members of the public during the intermediate phase of the incident.

NOTE - From EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Accidents (Table 2-2). EMERGENCY PLAN REVISION 76 PAGE 66 OF 192

J TABLE 6.4-2 INITIATION TIMES FOR PROTECTIVE ACTIONS FOR THE GENERAL PUBLIC APPROXIMATE INITIATION TIME EXPOSURE PATHWAY ACTION TO BE INITIATED Evacuation, in-house shelter, (if evacuation Is not practical), access O - 4 Hours Inhalation of gases or particulates control, respiratory protection, prophylaxis (thyroid protection). Direct radiation Evacuation, in-house shelter, access control. Take cows off pasture, prevent cows from drinking surface water, 4 - 48 Hours Milk quarantine contaminated milk, utilize stored feeds. Harvested fruits and vegetables Wash all produce or impound produce. Drinking water Cut off contaminated supplies, substitute from other sources. Unharvested produce Delay harvest until approved. 2 - 14 Days Harvested produce Substitute uncontaminated produce. Milk Discard or divert to stored products, such as cheese. Drinking water Filter, demineralize, test. EMERGENCY PLAN REVISION 76 PAGE 67 OF 192

7. EMERGENCY RESPONSE FACILITIES AND EQUIPMENT This section of the CNS Emergency Plan describes the Control Room and the Emergency Response Facilities, including on-site and off-site communications systems, assessment equipment and facilities, protective facilities and equipment, first aid and medical facilities, and damage control equipment and supplies.

7.1 CONTROL ROOM Emergency assessment and control is initially directed from the Control Room by the Shift Manager prior to activation of the Technical Support Center (TSC). The Control Room is located in the Control Building and is equipped with an Emergency Bypass Ventilation System allowing habitability during Design Basis Accidents as defined in the Cooper Nuclear Station (CNS) Updated Safety Analysis Report. The Control Room contains plant instrumentation, required technical drawings, CNS records, and communications equipment. Communication equipment available for this facility is shown on Table 7.1-1. 7.2 EMERGENCY RESPONSE FACILITIES When activated, the following Emergency Response Facilities supplement the Control Room in assessing and controlling emergencies: 7.2.1 TECHNICAL SUPPORT CENTER 0 The TSC (Figure 7.2-1) is the focal point for on-site emergency coordination and for directing and assisting the Control Room during station emergency conditions. The following functions are performed in the TSC:

Provide management and technical support to station Operations personnel during an emergency.
Relieve Operations personnel of duties not directly related to Reactor System manipulations (e.g., NRG notifications).
Temporarily assumes the role of the EOF in the event the EOF functions must be transferred.

The TSC is located on the first floor of the Administration Building. Figure 7.2-1 provides a layout of the TSC. If the TSC should become uninhabitable, select TSC personnel would relocate to the Control Room. Remaining TSC personnel would report to the AOSC. To protect personnel under most conditions, the TSC is provided with dedicated radiological protection and monitoring equipment capable of continuous indication of dose rates and airborne radioactivity concentrations. Local alarms provide early warning to TSC personnel. EMERGENCY PLAN REVISION 76 PAGE 68 OF 192

The TSC Ventilation System is comparable to the Control Room Ventilation System. Although not seismically qualified, redundant, or automatically activated, it does include high efficiency particulate air filters and charcoal filters. System capacity is based on design basis accident airborne radioactivity levels, independent of thyroid blocking provisions (potassium iodide). Emergency equipment listed in EPIP 5.7.21 and Appendix Eis provided to protect personnel who must exit the TSC or continue TSC operation during the presence of low-level airborne radioactivity or radioactive surface contamination. To facilitate the TSC function, a set of As-Built drawings of the station, schematics and diagrams, Technical Specifications, Station Operating Procedures, Emergency Operating Procedures, Severe Accident Guidelines and associated Technical Support Guidelines, station operating records, records needed to perform the functions of the EOF when it is not operational, and a copy of the Updated Safety Analysis Report are available to personnel in the TSC. Communication equipment available for this facility is shown on Table 7.1-1. 7.2.2 OPERATIONAL SUPPORT CENTER The OSC (Figure 7.2-2) is the assembly and staging area for CNS personnel for emergency response assignments. The OSC provides a location where plant logistic support can be coordinated during an emergency. Chemistry/Radiological Protection personnel, Mechanical Maintenance personnel, Electrical and Instrument & Control personnel, Utility personnel, and administrative personnel report to the OSC. The OSC is located adjacent to the TSC. In the event the OSC would need to be relocated due to habitability concerns, the Training Center Classrooms H, I, and J (903' Level) have been identified as an alternate OSC. Communication equipment available for this facility is shown on Table 7.1-1. 7.2.3 EMERGENCY OPERATIONS FACILITY The EOF (Figure 7.2-3) is located at 902 Central Avenue in Auburn, Nebraska approximately 11 miles west of the plant site. The EOF performs the following functions:

Management of the off-site emergency response.
Coordination of radiological and environmental assessment.
Determination of recommended protective actions for the public.
Coordination of emergency response activities with federal, state, and local agencies.
Event Classification.
Notification to off-site authorities.

EMERGENCY PLAN REVISION 76 PAGE 69 OF 192

The EOF has sufficient space to accommodate CNS emergency response personnel and representatives from local, state, and federal response agencies. Field Monitoring T earns will be dispatched from the OSC and controlled from the EOF. Emergency equipment in the Communications Building, West Warehouse, and/or Field Monitoring Team vehicles consist of material and equipment needed for off-site monitoring and re-entry activities. This equipment includes procedures, protective clothing, radiation detection instrumentation, dosimetry, air sampling equipment, respiratory protection equipment, personnel decontamination supplies, and counting instruments. A list of this equipment is included in the Emergency Plan, Appendix E, as well as in Implementing Procedure (EPIP) 5.7.21. Results of off-site surveys and sample analyses will be reported to the EOF for evaluation and assessment, and to aid in the development of protective action recommendations to off-site authorities. Personnel in the EOF and TSC have the capability to assess meteorological data, current p*lant conditions and release rate data from the Safety Parameter Display System (SPDS) or the Control Room to determine projected downwind doses. All of this data is prominently displayed in the EOF and is readily available to local, state, and federal authorities for use in making an independent determination of protective actions. Communication equipment available for this facility is shown on Table 7.1-1. Communication by facsimile equipment between the EOF, TSC, and the JIC is also provided. Information available in the EOF includes: CNS Technical Specifications, Operating Procedures, Emergency Operating Procedures, Updated Safety Analysis Report, environs radiological monitoring records, and selected As-Built drawings. In addition, copies of State and Local Emergency Response Plans and information pertinent to evacuation is also maintained. The Nebraska Emergency Management Agency's Mobile Operations Center (MOC) and the Nebraska State Patrol Mobile Command Post may be stationed adjacent to the EOF. These vehicles are self-sustaining with their own electrical power and communications systems. EMERGENCY PLAN REVISION 76 PAGE 70 OF 192

7.2.4 JOINT INFORMATION CENTER The JIC is the media briefing area and is the focal point for contact with the media. The JIC (Figure 7.2-4) is located adjacent to the EOF at 902 Central Avenue in Auburn, NE. The JIC is jointly staffed by utility, State, and Federal personnel. The JIC carries out the following functions:

Coordinate the dissemination of infonnation to the news media.
Conduct media monitoring.
Maintain rumor control.
Provide NPPD employees with infonnation concerning the emergency.

To assure accurate and timely infonnation is available to the public, personnel manning the JIC have current infonnation on plant status available. A direct continuous line for communication is available between the JIC and the other ERFs. NPPD personnel are available to respond to any questions regarding plant status, radiological releases, protective actions, etc. The JIC contains up-to-date copies of station, state, and county emergency plans, maps of the CNS site area and its environs, regional maps, and station layout drawings. Other equipment, facilities, and services that will be located within, or near the JIC include communication links with the EOF and state Emergency Operations Centers, reproduction equipment, and word processing capability.

O Communication equipment available for this facility is shown on Table 7 .1-1.

Since public infonnation activities occur at the JIC, CNS has not dedicated an area within the EOF for media representatives. The EOF has an area dedicated for state Public lnfonnation Officers to infonn counterparts located in the JIC of the events which are occurring at CNS. The NPPD Designated Spokesperson located in the JIC is responsible for all interfaces with the media. The Designated Spokesperson and the JIC Support Staff will receive information on plant status from designated personnel located in the EOF and will interact with state Public Information Officers as required. EMERGENCY PLAN REVISION 76 PAGE 71 OF 192

7.3 COMMUNICATIONS SYSTEMS AND NOTIFICATION 7.3.1 PLANT COMMUNICATIONS EQUIPMENT* On-site communications are provided by:

Site PBX.
EOF and JIC telephones, which are connected to the CNS PBX via NPPD owned fiber cable backed up by leased circuits.
CNS On-Site Cell Phone System.
Station lntercom/Gaitronics.
Alternate Intercom System.
FM Radio Systems with remote control consoles located in the Control Room and the Central Alarm Station, Secondary Alarm Station, TSC, OSC, and EOF (further described in Sections 7.3.4.1 and 7.3.4.2).

7.3.2 TELEPHONE COMMUNICATIONS Telephone communications to off-site organizations are provided by the following:

Trunks in buried cable connecting PBX to the central office at Brownville, Nebraska.
Federal Telecommunications System (FTS) Telephone System including 0 Emergency Notification System (ENS) and Health Physics Network (HPN) circuits. The ENS circuit of this system is manned 24-hours a day at the Control Room .;ind NRC Headquarters.
Trunks connecting the PBX to the N.P.P.D. microwave telephone network.
Several local numbers connecting telephones located in several places throughout the plant to the local service provider's Central Office in Brownville, Nebraska.
The State Notification Telephone System's dedicated lines will ring the Nebraska State Patrol, Missouri State Patrol, Atchison County 911 Center, Nemaha County Sheriffs Department, and the Richardson County Sheriff's Office. The use of law enforcement agencies and emergency services dispatch centers as the initial points of contact provides 24-hour coverage.

The dedicated lines listed also have extension lines in the following facilities: Nebraska State Emergency Management Agency EOG, Missouri State Emergency Management Agency EOC, Atchison County EOC, Nemaha County EOC and the Richardson County EOG. Once the EOCs are operational, notifications may be made using the extension lines at the EOCs with the concurrence between the respective EOG and the law enforcement and emergency services dispatch agencies. The CNS telephone system is split into two functional networks; one for normal business operations and one for emergency communications (CNS Emergency EMERGENCY PLAN REVISION 76 PAGE 72 OF 192

Preparedness (EP) telephones). Function of the CNS EP telephones is described below. Should the PBX lose AC power, the PBX and on-site Emergency EP analog telephones will automatically switch to back-up battery power. These batteries will power the PBX for approximately 6 hours. The Control Room digital telephones automatically switch to back-up battery power, and can also be powered from a site Emergency Diesel Generator. TSC and OSC digital telephones do not have battery back-up power, but can be powered from a site Emergency Diesel Generator. EOF and JIC telephone function is very similar to the telephones on-site at CNS. These telephones are connected to the CNS PBX via NPPD owned fiber cable backed up by leased circuits. Should there be a loss of AC power, a back-up source is available to power the telephone system. If the PBX connectivity were to be interrupted, network routers would go into a survivable mode, in which select telephones would have direct access to the local telephone service provider's Central Office (CO) lines. If the PBX System should fail, the system is designed to connect several extensions, designated as bypass telephones, directly to the local telephone service provider's CO lines. National Warning System (NAWAS) is installed, monitored, and operated from the Control Room. NAWAS, which is manned 24-hours a day, is a Nationwide Telephone System primarily for attack warnings. 7.3.3 SATELLITE TELEPHONES Should all installed means of off-site communications fail, CNS has remote satellite telephone extensions located in the Control Room, TSC, EOF, and JIC. Handheld satellite phones are also available for Fire Brigade and Field Monitoring Teams. 7 .3.4 RADIO COMMUNICATIONS For use both on-site and off-site are provided by: 7.3.4.1 TWO-WAY FM RADIO SYSTEM (BASE 1 AND BASE 2)

A system used primarily by in-plant Operations/Emergency Response personnel.
It is possible to operate this system either base-to-portable or portable-to-portable. The range of the Base 1 System at maximum would be approximately 20 miles.

7.3.4.2 TWO-WAY FM RADIO SYSTEM (LOW BAND)

This Low Band System is primarily used by off-site field monitoring personnel.
This Radio System can be used for communications between CNS and CNS EP vehicles.

EMERGENCY PLAN REVISION 76 PAGE 73 OF 192

At CNS ERFs three remote control heads are accessible, one each in the Control Room, OSC, and the EOF. Low band radio consoles are n

I / also located in CAS, SAS, and the 345 kV Substation. 7.3.4.3 COMMUNICATIONS WITH THE NEMAHA COUNTY SHERIFF'S DEPARTMENT

The CNS Two-Way Radio System can be used for communicating between the Control Room, Security Alarm Stations, OSC, and the EOF to the Nemaha County Sheriff's Department located in Auburn, Nebraska, on the Sheriff's Department frequency or State-wide radio frequencies.

7.3.4.4 CNS SECURITY DEPARTMENT COMMUNICATIONS

The CNS Security Department uses its own Two-Way Radio Systems operating in the 460 and 800 Mhz.

7.4 NOTIFICATION BY EMERGENCY CLASS Notification schemes detailed by specific emergency classification including notification of the general public are contained in EPIP 5.7.2, 5.7.6, 5.7.27, and 5.7.23. The four classes of emergency defined in Section 4 require varying degree and scope of emergency response. The emergency organization for a NOTIFICATION OF UNUSUAL EVENT classification consists of the normal shift personnel. Normally, no 0 further site emergency staff augmentation is required, although several members of Station Management, including Senior Management personnel are notified. Notification of responsible State and Local Governmental Agencies, and well as the NRG, will also be performed. In an ALERT classification, the TSC, EOF, and OSC will be activated. Notification of state, local, and NRG authorities, as well as Station Management and Staff will be initiated. SITE AREA EMERGENCY and GENERAL EMERGENCY classifications require complete activation of all Emergency Response Facilities, including State and Local Emergency Operations Centers (EOCs). The complete emergency response notification scheme, depicted in Figure 7.4-1, shall be initiated. EMERGENCY PLAN REVISION 76 PAGE 74 OF 192

7.5 ENVIRONMENT AL ASSESSMENT CAPABILITIES This section outlines the equipment available at CNS for the evaluation and assessment of emergency conditions. Some of this equipment is used in the initial evaluation and classification of the emergency as described in Section 4. Other equipment and capabilities described in this section are used in the assessment, mitigation, and subsequent analysis and monitoring of areas, equipment, and the environment. In some cases equipment may serve for both initial and continual assessment. 7.5.1 SEISMIC MONITOR A seismic event monitor and recorder is located in the Control Room. This instrument, in conjunction with assessment of equipment damage within the station, will be a primary factor in determining the emergency condition and classification as a result of a seismic occurrence. Detailed seismic information can be obtained from the Conservation and Survey Division of the University of Nebraska. Additional seismic information can be obtained from the U.S. Geological Survey office of Earthquake Studies in Golden, Colorado. 7.5.2 METEOROLOG ICAL MONITORING One meteorological monitoring site is located at a grade elevation of 889 feet above mean sea level. A 100-meter tower located approximately 2,655 feet from the northwest comer of the Reactor Building is used to gather the meteorological data. The 100-meter tower has two independent Meteorological 0 Monitoring Systems (A and B) and gathers data at three levels, 100 meters, 60 meters, and 10 meters. The redundant signal cables from the tower to the plant are located in either protected duct banks or directly buried in separated routings so that only one path can be interrupted at a time due to construction or other activities. Digital data is available through PMIS. The Meteorological Monitoring System is powered locally from the 12.5 kV ring bus and from MCC "L" in the event of a loss of off-site power condition. The system monitors and continuously records the following:

Wind speed and wind direction are measured continuously at all tower levels. Wind system components include sonic anemometers and heaters.

The range for wind direction is 0-360° +/- 3°. The range for wind speed is 0-112 mph +/- 0.336 mph accuracy up to 11 mph with accuracy increase to

                   +/- 2% above 11 mph.

The A and B Systems on the tower calculate 15-minute averages of Sigma Theta. Sigma Theta is calculated and updated every minute for the 15-minute time constant. The sigma theta values are then accessed by the Data Acquisition System.

u EMERGENCY PLAN REVISION 76 PAGE 75 OF 192

Air temperatures are monitored at all six locations on the tower. Each temperature system on the tower is comprised of a platinum RTD temperature probe and motorized aspirated radiation shield to monitor the temperature at the various levels. The estimation of atmospheric stability for the A and B Systems is then calculated based on the vertical temperature difference between the 100-meter, 60-meter, and 10-meter tower elevations. The range and accuracy are -58 to +122°F +/- 0.18°F not to exceed 0.18°F between tower vertical calibration points.
Dewpoint is calculated using output from an aspirated humidity sensor and ambient temperature measurement at the ten-meter level on the tower.
Precipitation is measured using an electrically heated rain and snow gauge located near the base of the tower. This gauge uses the tipping bucket method to measure all forms of precipitation. The instrument has an accuracy of+/- 1% from 1 to 3 inches/hour and 3% overall from O to 6 inches/hour.

Backup meteorological data may be obtained from the National Weather Service Office located in Valley, Nebraska, which offers projected windspeed, wind direction, and temperature up to the 10,000-foot level. Information can be obtained by telephone or NAWAS. 7.5.3 MISSOURI RIVER MONITORING CNS is located on the Missouri River, river mile 532.5. This river is the only water source in the area that has the potential for causing major flood damage. Operations personnel record the river level once per shift. Backup information concerning river level and flow is available from the U.S. Geological Survey Station in Nebraska City, Nebraska, via the U.S. Geological Survey in Council Bluffs, Iowa. EMERGENCY PLAN REVISION 76 PAGE 76 OF 192

7.5.4 RADIOLOGICAL MONITORS n CNS maintains a Radiological Protection Program and related radiological equipment in accordance with NRC Regulations, guidelines, and recommended practices. This program, including personnel, procedures, and equipment is periodically inspected by the NRC. The program provides manpower and equipment to evaluate, assess, and perform follow-up action for radiological emergencies. In addition to the equipment used in normal radiological controls, a special inventory of emergency equipment is maintained both on-site and off-site for emergency radiological response. State Radiological Health Departments, nearby nuclear plants, Federal Radiological Assistance Plan Teams, and private organizations such as the Institute of Nuclear Power Operations and General Electric may be requested to provide manpower and radiological monitoring equipment during an emergency situation. Equipment vendors may also be contacted to provide equipment needed in an emergency situation. The instrumentation listed below at the described locations is used to evaluate radiological conditions and assess the radiological risks. Tables 4.1-1 through 4.1-4 may require the use of indications from these sources in classifying emergencies. 7.5.4.1 EFFLUENT RELEASE POINT MONITORS Each effluent release point is continuously monitored for particulates, 0 radioiodines, and noble gases. The fixed filters and radioiodine absorbing charcoal cartridges may be changed and taken to the radiochemistry laboratory for analysis and release level determination. Noble gas samples may also be taken and analyzed, or release rate versus monitor readings may be used to determine release rates for each effluent point. The CNS effluent monitors are capable of monitoring releases during normal and accident conditions. EMERGENCY PLAN REVISION 76 PAGE 77 OF 192

7.5.5 MAIN STEAM LINE MONITORS The four main steam lines are continuously monitored for radiation level. This instrument system consists of four channels, which contain the following:

An ionization chamber, used to measure the steam radiation level as it leaves primary containment.
A logarithmic monitor with local readout located in the Control Room.
An annunciator for low/inop, high, and high/high radiation readout located in the Control Room.
A recorder is located in the Control Room and is provided to continuously monitor four channels. If radiation levels increase above the alarm setpoint, visual and audible alarms are actuated in the Control Room.

7.5.5.1 STEAM JET AIR EJECTOR MONITORS Off-gas radiation levels are continuously monitored by the Steam Jet Air Ejector Monitors. Each monitor system consists of an ionization detector with readouts and recorder in the Control Room. Calibration curves relate count rate and flow versus release rate at the Steam Jet Air Ejectors. This flow path is independently monitored at the Elevated Release Point by monitors described in Section 7 .5.4.1. Measurement is made upstream of the 30-minute holdup line. If the alarm setpoint is exceeded and the alarm does not clear within 15 minutes, the off-gas holdup line isolates and the reactor will be shutdown. 7.5.5.2 AREA RADIATION MONITORS Areas within CNS are monitored with permanently mounted Area Radiation Monitors. Many of these monitors have local level indications and local audible and visual alarms, including level indications and audible and visual alarm annunciation in the Control Room. Area Radiation Monitors are in locations that may be occupied by station personnel or where higher than normal radiation levels may indicate system malfunction. The alarm setpoints are based on the normal operational radiation background surrounding each individual monitor. Any abnormal increase in radiation levels will produce an annunciator alarm in the Control Room. A continuous record of this monitor system is provided on a Control Room multipoint recorder. Area Radiation Monitors located inside the primary containment will alert personnel to abnormal radiation increases. EMERGENCY PLAN REVISION 76 PAGE 78 OF 192

7.5.5.3 LIQUID RADIOACTIVITY MONITORS Changes in the levels of radioactive material within a liquid system with its subsequent radiation level change is normally monitored by the Area Radiation Monitor System. The Reactor Equipment Cooling System, the Radioactive Liquid Effluent Line, and the Service Water Cooling System each have detectors, which continuously monitor radioactivity level. A readout, visual and audible alarms, and recorder are provided in the Control Room. 7.5.5.4 CONSTANT AIR MONITORS In addition to the Gaseous Effluent Release Points described in Section 7.5.4.1. Constant Air Monitor units analyze surrounding building air. These units filter the air onto fixed particulate filters or iodine filters and monitors them for radioactivity level. Two of these monitors are normally stationed on the refueling floor, one monitoring for particulate radioactivity and the other monitoring for Iodine activity. These Constant Air Monitors are mobile and may be stationed at strategic locations within the station. Each has a local activity readout and visual alarm functions. 7.5.5.5 PORTABLE AIR SAMPLING EQUIPMENT In addition to the constant air monitors described above, portable air 0 sampling equipment is also available. Three basic types of portable air samplers are provided at the station. Particulate and iodine samplers, both high volume and low volume, may be used. These types of air samplers circulate air through a particulate filter and then through a radioiodine absorbing charcoal cartridge. The filters and charcoal cartridges are then analyzed for airborne radioactivity. Radioanalysis of these air samples are normally performed in the CNS Radiological Protection Counting Room or the Radiochemistry Laboratory. Both silver zeolite charcoal and conventional charcoal cartridges are available for use with portable air samplers. The use of conventional cartridges requires the extra step of purging noble gases prior to analysis for radioiodines. Facilities to purge these gases are available in the Radiochemistry Lab. Air samplers are also available in the emergency lockers described in EPIP 5.7.21. The Nebraska and Missouri State Radiological Health Departments also have air-sampling equipment available for off-site surveys. In addition, nearby nuclear plants, as well as the Department of Energy (DOE) via the NRG, may be requested to provide air-sampling backup. EMERGENCY PLAN REVISION 76 PAGE 79 OF 192

7.5.5.6 PROCESS MONITORS System readouts, scaled to provide normal and abnormal primary system parameter indications, are displayed in the Control Room. Such parameters include, but are not limited to, pressure, temperature, level indication, and flow rates. Each parameter will normally have alarm setpoints with audible and visual alarm capabilities when setpoints are exceeded. Particular instruments also have recorder capabilities, which serve as a record of system performance. Process monitors, their alarms, annunciators, and recorders, give indication of system performance. Process monitors serve as indicators to alert station personnel of emergency situations. Monitor indications, in conjunction with the Emergency Action Levels discussed in EPIP 5.7.1, are used to initiate emergency measures and aid in the evaluation and assessment of such emergency situations. 7.5.6 ENVIRONMENTAL RADIATION SURVEILLANCE The CNS Environmental Radiation Surveillance Program was initiated in 1970. The initial phase determined background levels of radioactivity in the environment around CNS. The Environmental Radiological Program has continued throughout plant startup, pre-operational testing phases, and operation of the station. This program basically consists of the aspects described below: 0

Radiation monitoring within a ten-mile radius, which is accomplished by the use of Dosimeters of Legal Record (DLRs). In addition, the NRC and State Health Departments have placed thermoluminescent dosimeters in areas surrounding the station. -
Continuous air sampling is performed within a ten-mile radius of the station.

Air sampling stations collect airborne particulates on stationary filters, then pass this filtered air through charcoal cartridges to collect ,gaseous radioiodines.

Samples which are periodically taken for e'nvironmental radiological assessments include:
Missouri River water samples.
Ground water samples.
Vegetation samples.
Milk samples.
Fish samples from Missouri River.
Sediment from Missouri River shoreline.

u EMERGENCY PLAN REVISION 76 PAGE 80 OF 192

The environmental program analyzes a wide variety of samples from the environment for radiological concerns. In case of an emergency affecting off-site radiological parameters, any phase of this program may be used to assess the situation. In addition, portable air sampling equipment for particulates and radioiodines is available for gaseous release emergencies; dose rate instruments and analytical capabilities are available for environmental assessment. The State(s) Department of Health, other nuclear plants, and the DOE via the NRG are also available to aid in environmental assessments. 7.5.7 RADIOANAL YSIS LABORATORIES The radiochemistry laboratory and Radiological Protection Counting Room, located in separate buildings from the reactor, are used to analyze on-srte and off-site emergency monitoring samples. These facilities are equipped with analytical instruments capable of measuring radionuclides and their associated emergency concentrations in environmental samples. Gross radiation level instrumentation is also available in the EOF. The State(s) Health Department and Omaha Public Power District's Fort Calhoun Nuclear Station may also be requested to assist in the analysis of environmental monitoring samples. 7.6 FIRE PROTECTIO N Although the Fire Protection System is not detailed in the CNS Emergency Plan, the type, location, and severity of a fire may prompt this Emergency Plan to be implemented. CNS maintains a Fire Protection System in accordance 0 with NRC regulations, which is *periodically inspected by the NRG to verify fire protection capabilities. CNS has its own fire pumps with backup capabilities and distribution systems, including numerous hose and hydrant stations, automatic sprinkler system, and fire barriers. Materials used meet fire code requirements. An on-site emergency fire house containing portable fire fighting equipment is maintained and periodically inventoried. Smoke detectors, heat detectors, visual means, and Control Room annunciation of the fire system serve as indicators of fire location and severity. 7.7 EMERGENC Y LOCKERS Emergency lockers containing respiratory protection equipment, air sampling equipment, survey equipment, and other emergency type equipment are located in the Control Room and the CNS Emergency Response Facilities. An emergency rescue locker is located in Administration Building, level 903' near the entrance to the RCA. EPIP 5.7.21 lists the locations and minimum inventory requirements for the emergency lockers, the station ambulance, and the Field Monitoring Team vehicles. EMERGENCY PLAN REVISION 76 PAGE 81 OF 192

Respiratory protection equipment, protective clothing, survey equipment, sampling equipment, and other equipment for re-entry, rescue, or emergency n operations will be provided from these emergency lockers if normal station supplies are not available. Radioprotective tablets (Potassium Iodide) are stocked in emergency lockers and are available on a voluntary basis to emergency response personnel as conditions dictate. These tablets will be distributed only with the permission of the Emergency Director. These lockers will remain unlocked with a seal across the door, which will break when the doors are opened. The contents of the lockers will be inventoried at least each calendar quarter and also after each use using EPIP 5. 7.21. Any missing or expended items shall be replaced. A new seal will then be attached to the doors. Any time a locker seal is found broken, the contents of the locker will be promptly verified by inventory of the contents. The procedures and associated attachments available in the emergency lockers will be updated as required. 7.8 HABITABILITY EQUIPMENT Control Room shielding and ventilation are designed to allow personnel habitability during Design Basis Accident Conditions. The TSC/OSC has shielding and ventilation similar to the Control Room for habitability during the course of an emergency. The TSC/OSC Ventilation System is not seismic 0 qualified, redundant, instrumented in the Control Room, or automatically activated. In the unlikely event the OSC is not habitable, an Alternate OSC location is provided which has ventilation similar to the TSC/OSC. The EOF meets the habitability requirements of NUREG-0696 (Functional Criteria for Emergency Response Facilities). Portable radiation monitoring instrumentation, communications equipment, respiratory equipment, and protective clothing are available in or near the Control Room, TSC, OSC, and AOSC. Portable radiation monitoring instrumentation, communications equipment, and protective clothing are available in or near the EOF. EMERGENCY PLAN REVISION 76 PAGE 82 OF 192

7.9 MEDICAL FACILITIES AND FIRST AID n 7.9.1 MEDICAL FACILITIES Arrangements have been made with the hospitals listed below for care of injured personnel, including cases involving radiological contamination and radiation over-exposure. Selection of the hospital and medical assistance will normally be based on obtaining the most rapid access to the necessary medical services, the capability of the specific hospital to provide the required services, and the preference of the injured personnel, if the type and severity of the injury permit.

Nemaha County Hospital, 2022 13th Street, Auburn, Nebraska.
University Nebraska Medical Center, Hixson-Lied Center for Clinical Excellence, East Side, 44th and Dewey, Omaha, Nebraska.

7.9.2 FIRST AID First aid kits are available at strategic locations throughout the station. These kits are fully equipped with supplies and materials appropriate for use in radiological emergencies. 0 EMERGENCY PLAN REVISION 76 PAGE 83 OF 192

Figure 7.2-1 TSC Floor Plan

                                            ~
                                             ~

rn

?~

D UkitJ I EMERGENCY PLAN REVISION 76 PAGE 84 OF 192 I

Figure 7.2-2 OSC Floor Plan rn Tectli-cal wnur I I*:t , Opmtiaeal Ct!Wlt!C) EMERGENCY PLAN REVISION 76 PAGE 85 OF 192 I

Figure 7.2-3 EOF Floor Plan

            ~              C£:HE'RATOR
                      ~

NRC (:OJ:fFERENCE ROOM Tfll<X>M ROOM

1 17
1.21 YIOE1>

CONF'E.R£1,ICE ROOM L DOSE ,U SESSJ,iENT P.OOM 11 ~ *

RllhP ROOM

                                                  '°'

MlSSOUR l Scolo: 1 ;a* ., , 1* Cl'OD Fi LE CC05.J539 EMERGENCY PLAN REVI SION 76 P AGE 86 OF 192

Figure 7.2-4 JIG Floor Plan

                          °""""'° .,,,_"'"

AOOM 127 JIC 117 """" 119 AOOW 121 WlCO ROOM 11 6 ffl ) """" 112 115 ROOM ROOW 101

                                                                     >JJOITORllM ROOM 103 MEDIA ROOM Scale: I/ 8 *

E MERGENCY PLAN REVI SION 76 PAGE 87 OF 192

Figure 7.4-1 Notification Chart for Emergency Classification Shift Manager or Emergt>ncy D11ector Shift Communicator Nebnsk3 State Patrol M15:i0Ufl State Pat, ol L Nebr~ska Emergenc.y M~nageme11t /,fency L Missouri State Emergency MJnagement Agen r~emahd Cu1,nty Atch1,on County 911 Sheritt Center L Nemaha Count*, Emer,;ency Management L AtChlS0n County Emergency Management ) R1chMdson County NRC Resident Sheriff Inspector L Richardson County Eme1gency Management Public Affairs Duty ERO (By Procedure) Officer NRC ENS EMERGENCY PLAN REVISION 76 PAGE 88 OF 192

Table 7.1-1 ERF Communications Systems COMMUNICATIONS SYSTEM osc EOF TSC CR JIC AOSC

1. Telephone PBX X X X X X X

2. Station Intercom System X X X X "Gaitronics"

3. Alternate Intercom System X X X X X X

4. FTS 2001 SYSTEM X X X

ENS Telephones to NRC
HPN Telephones to NRC X X
RSCL Telephones to NRC X X X
MCL Telephones to NRC X
PMCL Telephones to NRC X X

5. NPPD Microwave Network X X X X X X

6.

Telephone extensions to Local X X X X Exchange

7. NAWAS X 0 8. CNS State Notification Telephone System X X X

9. Radio Base Station Console X X X X (Base 1/Base 2)

10. Nemaha County Sheriff's Dept.

X X X Radio

11. NPPD State-Wide Radio System X X X (Low Band)

12. CNS On-Site Cell Phone System X X X X

13. Satellite Telephones X X X X EMERGENCY PLAN REVISION 76 PAGE 89 OF 192 I

8. MAINTAINING EMERGENCY PREPAREDNESS Maintenance of the CNS Emergency Plan and Emergency Preparedness Program consists of: (1) training for NPPD emergency response personnel, (2) drills and exercises, (3) regular emergency plan review and evaluation, and (4) periodic inventory, maintenance, and testing of emergency facilities and equipment.

8.1 TRAINING Emergency Preparedness training ensures that Emergency Response Organization (ERO) members will be familiar with applicable portions of the following Emergency Preparedness documents:

NPPD Emergency Plan for CNS.
NPPD Emergency Preparedness Implementing Procedures (EPIPs) for CNS.
ERO Positional Instructional Manual Checklists (PIMs) for CNS.

8.1.1 TRAINING FOR CNS EMERGENCY RESPONSE ORGANIZATION (ERO)

               ~ ERO members may be stationed at CNS, Columbus General Office, or other NPPD offices. These employees may become members of the ERO by virtue of their normal job position, or may be selected to fill a position in the ERO based on their availability and/or personal qualifications.

Emergency Response Organization members will receive initial training which 0 will be followed by requalification training. This training includes both knowledge-based and performance-based elements. Requalification training on knowledge based elements generally will be offered on a 12-month cycle not to exceed 15 months. Requalification training on performance-based elements will be offered on a calendar year basis. Exceptions to the requalification training periods are as noted in ERO Training Qualification Descriptions. Details are as noted in the ERO Training Qualification Descriptions. No employee will become an active member of the ERO until training has been completed and the individual has been qualified. Training will be conducted in accordance with the ERO Training Program Procedure. Initial training will consist of formal sessions utilizing materials indicated by the ERO Training Program Procedure. This training will be followed by an evaluation of the student's comprehension of the subject material. Requalification training may consist of formal training, drills, exercises, or other alternate methods of completion as described by the ERO Training Program Procedure. Training records and documentation will be maintained by the Nuclear Training Department. The Emergency Preparedness Department is responsible for assuring that all appropriate emergency response personnel are adequately trained. EMERGENCY PLAN REVISION 76 PAGE 90 OF 192

Training will be developed and utilized for areas, as required, per 10CFR50 Appendix E.IV.F, Training:

Directors and/or coordinators of the plant emergency organization.
Personnel responsible for accident assessment, including Control Room shift personnel.
Radiological Monitoring Teams.
Fire Control Teams (fire brigades).
Repair and Damage Control Teams.
First Aid and Rescue Teams.
Medical support personnel.
Licensee's headquarters support personnel.
Security personnel.

Emergency Preparedness training material is identified in the ERO Training Program Procedure. A listing of training requirements and a synopsis of the course content is contained in this document. Fire Brigade, Security, and First Aid training have been established to fulfill requirements of other programs. Lesson plans addressing these areas have been developed and are taught under their respective training programs. Training materials will be revised to correspond with changes made to the Emergency Plan, Emergency Plan 0 Implementing Procedures, or other supporting documents. Changes may be identified through drill and exercise performance. 8.1.2 TRAINING FOR EMERGENCY PREPAREDNESS DEPARTMENT PERSONNEL Training for the Emergency Preparedness Manager and staff will be provided through participation in industry sponsored emergency planning symposia and workshops, as well as observation of drills and exercises of other utilities. This training will also be conducted on a 12-month cycle not to exceed 15 months. Documentation for such participation will be recorded and maintained by the CNS Training Department. 8.1.3 TRAINING FOR PARTICIPATING AGENCIES Training for participating agencies is programmed by the individual agencies with aid from the State Governments in Nebraska, Missouri, Kansas, and Iowa. NPPD personnel are available to describe the special conditions and constraints involved in dealing with the station emergencies and any radiological release situations. NPPD offers tr:aining annually for employees of the Nemaha County Hospital, Members of the Volunteer Fire Departments of Brownville, Nemaha, Peru, and Auburn, Local Ambulance Services, Local Emergency Management Personnel, and Local Law Enforcement Agencies. EMERGENCY PLAN REVISION 76 PAGE 91 OF 192

This training includes notification procedures, basic radiation protection theory, and the identity, by position and title, of the individual in the on-site emergency organization who will control CNS emergency response activities. 8.1.4 PUBLIC EDUCATION NPPD prepares educational material for annual distribution to the public within the 10-mile plume exposure Emergency Planning Zone (EPZ). The material is mailed to residents and is available for review at NPPD headquarters, CNS, the Nebraska State Emergency Management Agency, the Nebraska Department of Health - Division of Regulation and Licensure, and the Missouri State Emergency Management Agency. The material outlines the station's operational concept, defines the various classifications of emergencies, summarizes the emergency plan and procedures developed to safeguard the general public, reviews appropriate protective actions (i.e., in-house shelter, evacuation, etc.), and describes public warning signals and their meaning. Facts about radiation and contacts for additional information are included. The material is reviewed annually by NPPD and State and Local Emergency Management Agencies and updated as required. To provide for the notification and education of the transient population within the 10-mile EPZ, NPPD has provided numerous copies of the information to the following:

Missouri Tourist Information Center.

0

Rock Port Rivers Edge Campground.

All motels within the 10-mile EPZ.

Indian Cave State Park.
Brownville Historical Society.

In the event of an emergency at CNS, the owner, operator, etc., of each of these establishments has been instructed to distribute the material to any individuals occupying their facilities. For special use areas within the EPZ, NPPD has made arrangements for the e$tablishment of other means of notification of the transient population. Informational bulletin boards and/or signs have been established at Brownville recreation area, Brick Yard Hill Wildlife Area, Steamboat Trace Trail and river ' access boat ramps. At Indian Cave State Park an informational brochure includes information on what actions should be taken in the event of an accident at CNS. 8.1.5 MEDIA FAMILIARIZATION Annual programs are conducted to acquaint media personnel with the CNS Emergency Plan, information concerning basic nuclear plant operation and radiation, and the locations and means employed to disseminate emergency information to the public. This activity may be performed in cooperation with the NPPD Corporate Communications Department. EMERGENCY PLAN REVISION 76 PAGE 92 OF 192

8.2 DRILLS AND EXERCISES Regular participation by station personnel in drills and exercises is designed to maintain emergency preparedness and test specific aspects of emergency plans, procedures, and equipment. Evaluation of these drills and exercises is conducted and revisions to the Emergency Preparedness Program are implemented to improve performance. 8.2.1 Two drills or exercises, one of which shall be unannounced, will commence off-hours between 6:00 pm and 4:00 am during each eight-year cycle. These drills or exercises will be conducted during different seasons of the year. Scenarios used shall provide reasonable assurance that anticipatory responses will not result from preconditioning of participants. Scenarios must include a wide spectrum of radiological releases and events, including hostile action. Exercise and drill scenarios as appropriate must emphasize coordination among onsite and offsite response organizations. 8.2.2 EXERCISES An exercise is an event that tests the integrated capability and a major portion of the basic elements existing within the Emergency Plan. Emergency exercises (required biennially) are conducted annually and simulate events that may result in off-site radiological releases to the extent requiring response by off-site authorities. In accordance with the Nuclear Regulatory 0 Commission (NRG) and the Federal Emergency Management Agency (FEMA) rules, these exercises are conducted jointly with participating federal, state, and local government agencies to assure effective response to major emergency situations. These combined exercises are coordinated by the CNS Emergency Preparedness Department with State and Local Emergency Planning Personnel. Objectives for joint exercises are developed by NPPD with the states of Nebraska, Missouri, Kansas, and Iowa as appropriate. Scenarios shall vary from exercise to exercise and are developed to ensure that essential portions of plans and organizations are tested within an eight-year cycle and to ensure that the scenarios include at least the following:

Statement of basic objectives and evaluation criteria.
Date, time, place, and participating organizations.
Simulated events.
Time schedule of initiating events.
Narrative summary describing exercise particulars which may include such things as simulated casualties, off-site medical assistance, rescue of EMERGENCY PL.AN REVISION 76 PAGE 93 OF 192

personnel, deployment of Radiological Monitoring Teams, and public information activities. n

Provision for proper utilization of observers.

Initial critiques will be held at each facility immediately following termination of the exercise. A synopsis of the exercise critique process is as follows: Players critique their own performance, noting areas requiring improvement. Evaluators then present their initial findings to the players. Players have the opportunity to provide evidence to the Evaluators of actions taken for which they did not receive credit. Evaluators then prepare their critique findings. This includes categorizing each finding using the following definitions. Deficiency Demonstrated or observed inadequacies, whether a single isolated case or a collection of observations, that indicate the state of emergency preparedness is not adequate to protect the health and safety of the public. Weakness Demonstrated or observed inadequacies, that require corrective action, but when considered by themselves do not adversely impact the health and safety of the public. 0 Improvement Items Demonstrated or observed problem areas that are not consider~d to adversely affect the health and safety of the public, but correction would enhance the level of preparedness. Evaluators should help identify the root cause and, if possible, provide a solution for deficiencies. Evaluators should also help provide insights and/or solutions for weaknesses. The Emergency Preparedness Department conducts a meeting of the Lead Evaluators where categorized critique items are formatted for presentation to CNS Sr. Management. Exercise findings are then presented to the NRC. The Emergency Preparedness Manager will ensure that Exercise findings are tracked and resolved as appropriate. At the conclusion of joint exercises, FEMA, NRC, and State Observers will also conduct critiques. Formal evaluations of these exercises published by federal or state authorities will be reviewed by NPPD Management. Areas found to be deficient or we(:lk will be identified and corrective actions implemented. The Safety Review and Audit Board, as well as the NPPD Quality Assurance Department, will perform periodic audits of the Emergency Preparedness Program and may serve as observers during exercises. EMERGENCY PLAN REVISION 76 PAGE 94 OF 192

8.2.3 DRILLS A drill is a supervised instruction period aimed at testing, developing, and maintaining skills in a particular operation. Emergency drills are conducted on a scheduled basis with emphasis placed upon the orderly implementation of activities prescribed within the Emergency Plan and EPIPs. Guidelines for administering drills are in place to ensure a quality drill program. Drill performance is critiqued by personnel acting as drill Evaluators who may offer on-the-spot corrections to deficient performance. Each Evaluator is assigned to evaluate drill performance in a specific area of emergency response. A written evaluation of drill performance is provided to CNS Management by the Emergency Preparedness Manager. Based on the results of these critiques, including participants' comments, follow-up action is then recommended by the Emergency Preparedness Manager, with action items assigned by the appropriate level of management. Drills for the station staff are conducted periodically to: (1) test response timing and familiarity with implementing procedures and methods, (2) test emergency equipment, (3) ensure that emergency response organization personnel are familiar with their duties. Certain drills (i.e., fire, communications and notification, and medical emergency) are coordinated with off-site participating 0 agencies. The Emergency Preparedness Manager has the overall responsibility for preparing, scheduling, and conducting emergency drills. A brief description of the type of drills conducted at CNS follows: 8.2.3.1 FIRE DRILLS Fire drills are conducted in accordance with plant Technical Specifications. 8.2.3.2 MEDICAL EMERGENCY DRILLS A medical emergency drill or exercise involving the treatment of a simulated contaminated person is conducted once per calendar year with provision for participation by local support service agencies. 8.2.3.3 RADIOLOGICAL MONITORING DRILLS These drills are conducted annually for on-site and off-site personnel assigned to radiation survey and Field Monitoring Teams. They shall include operation of instruments, tests of field communications equipment, interpretation of radiation readings, calculation of dose rates, collection of sample media (soil, water, vegetation, and air) and record keeping. EMERGENCY PLAN REVISION 76 PAGE 95 OF 192

8.2.3.4 RADIOLOGICAL PROTECTION DRILLS n Radiological Protection drills are conducted semi-annually, generally in connection with joint exercises or radiological monitoring drills. They involve analyses of simulated elevated radiation levels, both liquid and airborne, as well as direct radiation measurements in the environment. Analyses of in-plant liquid samples including the use of the Post-Accident Sampling System will be conducted on an annual basis. 8.2.3.5 COMMUNICATIONS DRILLS Communications systems are periodically tested during normal use, CNS security checks, and scheduled tests, as well as during emergency drills and exercises. The CNS telephone System, Microwave Communications System, Site Radio System, and Plant Intercom System are used daily during normal plant operation. The NRC ENS is tested daily by phone check from NRC Headquarters and monthly from CNS to NRC Headquarters from the Control Room, Technical Support Center (TSC), and Emergency Operations Facility (EOF). The NRC Health Physics Network (HPN) direct telephone is tested monthly from the TSC and EOF. The State Notification Telephone System is tested monthly. 8.3 EMERGENCY PREPAREDNESS DEPARTMENT 0 To ensure the maintenance and implementation of the Emergency Preparedness Program, several Emergency Preparedness positions at CNS have been established. A CNS Emergency Preparedness Manager reports directly to the Director of Nuclear Safety Assurance. Emergency Preparedness Coordinator (EPC) positions have been established with each assigned a primary area of responsibility. Each EPC reports directly to the Emergency Preparedness Manager. An Emergency Preparedness Specialist (EPS) position has also been established to assist in the conduct of Emergency Preparedness Department Activities. The EPS reports directly to the Emergency Preparedness Manager. A Nuclear Instructor - Emergency Preparedness (NI-EP) position has been established with responsibilities to implement requirements of the Emergency Preparedness Training Program. The NI-EP also reports directly to the Emergency Preparedness Manager. All of the above positions have been established to:

Maintain continued coordination with State and Local Emergency Planners on the status of Emergency Preparedness including budgetary issues.
Annual review and development of revisions to the CNS Emergency Plan and Implementing Procedures.

u

Coordinate drills and other aspects of the NPPD Emergency Preparedness Training Programs.
Coordinate the development and implementation of the annual exercise.

EMERGENCY PLAN REVISION 76 PAGE 96 OF 192

Ensure that adequate District resources are available to support the NPPD Emergency Preparedness program.

' I

Assist the EP Manager with the following:
All emergency preparedness activities at the CNS site.
Emergency preparedness and public relations with the local communities; includes medical facilities.
Alert and Notification System including fixed siren maintenance and testing and Emergency Alert System radio distribution and maintenance.
Emergency Response Facility readiness.
Implement requirements of the EP Training Program including development and revision of CNS EP Training Lesson Plans, and maintenance of EP training records.
Responsibilities as defined in the Emergency Plan and EPIPs.
NRG and FEMA interface.
Audits; includes Quality Assurance audits per 10CFR50.54(t) and NRG Inspection Report response submittal.
Meteorological and dose assessment.
Annual exercise; includes scenario preparation.
Emergency Preparedness Drills.

0

Responsibilities as defined in the Emergency Plan and EPIPs.
NRG and FEMA interface functions relating to the above areas.

8.4 CORPORATE COMMUNICATIONS DEPARTMENT The CNS Emergency Preparedness Program relies on support from the NPPD Corporate Communications Department. Personnel within this department are responsible for:

Interface with the CNS Emergency Preparedness Department.
Interface with other agency Public Information Officers.
Emergency planning public information material to include mailing list control revision, reprint, and mailout of the material.
Media interfaces which may include conferences, quarterly newspaper and radio ads for the CNS Alert and Notification System testing program, media monitoring, and JIG activities.
Training in the above areas.
Responsibilities as defined in the Emergency Plan and EPIPs.

EMERGENCY PLAN REVISION 76 PAGE 97 OF 192

8.5 REVIEW AND UPDATE OF THE EMERGENCY PLAN The CNS Emergency Plan is reviewed annually and revised, as necessary. The annual review and update is documented by the CNS Emergency Preparedness Manager. Special attention is devoted to reviewing station-government agency interfaces, maintaining effective communication channels, and, on a quarterly basis, ensuring the accuracy of the contact and notification lists to verify telephone numbers and the responsible individuals to be contacted. liaison with State and Local Agencies ensures untform updating and plan improvement. All revisions to the Emergency Plan shall be reviewed and approved by the Station Operations Review Committee prior to implementation. A review of EALs shall be conducted by NPPD with State and Local Emergency Management Personnel on an annual basis. Independent audits of the various aspects of the Emergency Preparedness Program are conducted at least biennially by NPPD Quality Assurance Personnel per 10CFR50.54t and the results of such audits are reported to the Safety Review and Audit Board (SRAB). The independent audit includes, but is not limited to, the Emergency Plan, Emergency Plan Implementing Procedures, training, readiness testing, equipment, and interfaces with State and Local Organizations. The results are considered by NPPD Management in modifying aspects of the plan. Audit documentation is maintained for at least five years. Revised or updated emergency plans and procedures are handled in accordance with document control procedures as delineated in the CNS Operations Manual. Distribution is controlled by the 0 Emergency Preparedness Manager, via the Document Control Department. 8.6 MAINTENANCE AND INVENTORY OF EMERGENCY EQUIPMENT AND SUPPLIES Quarterly inspections of the operational readiness of items of emergency equipment and supplies are conducted. Deficiencies noted during inspections are corrected in a timely manner. The use of EPIP 5.7.21, Maintaining Emergency Preparedness - Emergency Exercises, Drills, Tests, and Evaluations, in conjunction with the CNS Preventative Maintenance Tracking System and Emergency Preparedness Departmental Guides ensure equipment is ready for use. Sufficient reserves of instruments and equipment are maintained to replace those undergoing calibration or repair. Calibration of equipment is conducted at intervals set forth in Technical Specifications. In addition, the planned use of communications, first aid, fire fighting and radiation measuring equipment during scheduled drills further ensures the availability and operability of emergency equipment. EMERGENCY PLAN REVISION 76 PAGE 98 OF 192

9. RECOVERY n This section of the Emergency Plan describes the initiating conditions and transitional steps required to move from Emergency Response Organization operations into Recovery Operations. With the safety of the public and station personnel being of the utmost priority, recovery operations allow for a smooth transition from the Emergency Response Organization operations to normal day-to-day operations.

Recovery operations will include measures taken during and immediately following the emergency, as well as the longer term post-emergency efforts. These operations will be performed by station and other NPPD personnel, and if required, by contract technical and labor support. Manpower and equipment resources supporting the individual functional segments of the Recovery Organization will vary according to the severity of damage and specific situational needs. The Emergency Director will evaluate the effectiveness of corrective actions and determine if the emergency is under control. The following station conditions will serve as general guidelines for decisions whether the emergency is under control:

Radiation levels are stable or decreasing with time.
Releases of radioactive materials to the environment have ceased or are controlled within permissible license limits.
Fire, flooding, or similar emergency conditions no longer constitute a hazard to the station or station personnel.

Q

Measures have been successfully instituted to correct or compensate for malfunctioning equipment.

Based on the consideration of these criteria as well as other pertinent items, the Emergency Director shall determine when to activate the Recovery Panel. 9.1 RECOVERY PANEL If after evaluating the effectiveness of corrective actions, the Emergency Director determines that the emergency is under control, he will activate the Recovery Panel. The panel may consist of the following personnel:

Emergency Director.
VP-Nuclear/CNO or his designee.
Emergency Operations Facility (EOF) Director.
Technical Support Center (TSC) Director.
JIC Director (JIC).
Radiological Control Manager (EOF).

\ __) EMERGENCY PLAN REVISION 76 PAGE 99 OF 192

Personnel acting on this panel can either be physically present or connected by telephone conference from their various Emergenc y Response Facilities. The purpose of the Recovery Panel is to evaluate emergenc y termination considerations, determine plant status parameters, and the planning and implementation of recovery operations. A time frame can then be established for securing emergenc y response and de-escalation of the emergency classification for recovery, if not already completed. Based upon this information, the VP-Nuclear/CNO or his designee may initiate the Recovery Organization. 9.2 RECOVER Y ORGANIZATION Once the decision has been made to activate the Recovery Organization, NPPD Emergenc y Response Facilities and their personnel shall be informed (via briefings, Public Address systems, telephones, etc.). The JIC Director shall inform personnel at the JIC of the activation of the Recovery Organization. Depending upon the amount of media interest, the JIC may be deactivated at this time. Public Affairs duties and responsibilities would then be assumed by the Corporate Communications & Public Relations Manager and Staff. Other personnel at the JIC return to their normal work stations and support the recovery effort through their normal position functions. The Emergenc y Director or his alternate shall inform personnel in the EOF, TSC, OSC, and Control Room of the activation of the Recovery Organization. Depending upon current conditions, any of these facilities may be deactivated at this 0 time. Once deactivated, personnel in these facilities would return to their normal work stations and support the recovery effort through their normal position functions. The Recovery Organization is the same as the normal Nuclear Power Group Organization described in Section 5, except the Corporate Communications & Public Relations Manager and the Corporate Environmental Manager are included in the organization. 9.3 RECOVER Y EXPOSUR E CONTROL The General Manager of Plant Operations is responsible for ensuring that the Radiation Protection Manager evaluates the advisability of initiating re-entry. Information on existing conditions, interviews with employees involved in the emergency, regulatory dose guidelines, and when necessary, counsel from recognized experts will be utilized in formulating decisions on re-entry. The developm ent and evaluation of these operations will be under the direction of the General Manager of Plant Operations. General Office support personnel will aid in the requisition of technical assistance, increased manpower, and special equipment. During recovery and re-entry operations, actions will be pre-planned to minimize the amount of radiation dose to personnel. Access to areas will be controlled and radiation dose will be documented. Estimates of total population dose will be coordinated with state and federal authorities. IEMERGENCY PLAN REVISION 76 PAGE 100 OF 192 I

9.4 NUCLEAR SAFETY COMMITTEES Normal safety review organizations, Station Operations and Review Committee (SORG), and Safety Review and Audit Board (SRAB) will continue to function throughout all recovery activities. 9.4.1 STATION OPERATIONS REVIEW COMMITTEE (SORG) An on-site nuclear safety committee is provided to review all matters pertaining to nuclear safety in the operation of the nuclear facility. This committee is advisory to the General Manager of Plant Operations and the chairman is designated in writing by the General Manager of Plant Operations. Commitfee membership, responsibilities, and authorities are detailed in Station Procedures. 9.4.2 SAFETY REVIEW AND AUDIT BOARD (SRAB) An off-site nuclear Safety Committee is provided to perform independent review and audit of station activities. SRAB is advisory in nature. Committee membership, responsibilities, and authorities are detailed in the SRAB Charter. IEMERGENCY PLAN REVISION 76 PAGE 101 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN APPENDIX A PROCEDURES IMPLEMENT1NG THE EMERGENCY Pl.AN APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN PROCEDURE NO. TITLE 5.7.1 Emergency Classification 5.7.1.1 Emergency Classification Process 5.7.2 Emergency Director EPIP Incident Command Post (ICP) Hostile Action Based Event Roles and 5.7.3 Responsibilities 5.7.6 Notification 5.7.7 Activation of TSC 5.7.8 Activation of OSC 5.7.8.1 Activation of Alternate OSC 5.7.8.2 Activation of Alternate Off-Site OSC!fSC 5.7.9 Activation of EOF 5.7.10 Personnel Assembly and Accountability 5.7.11 Early Dismissal/Evacuation of Site Personnel 5.7.12 Emergency Radiation Exposure Control 0 5.7.13 5.7.14 Personnel Monitoring and Decontamination Stable Iodine-Thyroid Blocking (Kl) 5.7.15 OSC Team Dispatch 5.7.16 Release Rate Determination 5.7.17 Dose Assessment 5.7.17.1 Dose Assessment (Manual) 5.7.18 Off-site and Site Boundary Monitoring 5.7.19 On-site Radiological Monitoring 5.7.20 Protective Action Recommendations Maintaining Emergency Preparedness - Emergency Exercises, Drills, 5.7.21 Tests, and Evaluations 5.7.23 Activation of the JIC 5.7.24 Medical Emergency 5.7.25 Recovery Operations 5.7.26 Long-Term Environmental Monitoring 5.7.27 Alert and Notification System 5.7.27.1 NOAA/EAS Radio Malfunction 5.7.27.2 False Activation of Alert and Notification System 5.7.28 Administration of Positional Instruction Manuals (PIMs) 5.7COMMUN Communications 5.7ENS ENS Communicator IEMERGENCY PLAN REVISION 76 PAGE 102 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN APPENDtXA PROCEDURES IMPLEMENTING THE EMERGENCY PLAN PROCEDURE NO. TITLE 0-EP-01 Emergency Response Organization Responsibilities Configuration Control of the Automated Notification System 0-EP-02 (ANS) 0-EN-EP-306 Drills and Exercises 0-EN-EP-308 Emergency Planning Critiques TPP 101 Emergency Response Organization TPP 102 Emergency Preparedness Staff Training and Qualification EPDG 2 Attachment H-1 CNS Drill and Exercise Manual - Scheduling EPDG 2 Attachment H-2 CNS Drill and Exercise Manual - Scenario Development CNS Drill and Exercise Manual - Drill and Exercise EPDG 2 Attachment H-4 Preparations EPDG 2 Attachment H-5 CNS Drill and Exercise Manual - Critique Process 0 IEMERGENCY PLAN REVISION 76 PAGE 103 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN ,.,-...) Controlled copies of the CNS EPIPs are maintained in all Emergency Response Facilities. Summaries of each EPIP and a cross-reference to the appropriate section of the CNS Emergency Plan is provided below. 5.7.1 EMERGENCY CLASSIFICATION This procedure provides a means of classifying an event into one of four emergency classifications as described in Section 4 of the Emergency Plan. An EAL is a pre-determined, site specific, observable threshold for a plant Initiating Condition (IC) that places the plant in a given Emergency Classification Level (EGL). An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (on-site or off-site); a discrete, observable event; results of analyses; entry into specific Emergency Operating Procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency classification level. EALs are utilized to classify emergency conditions. To the extent possible, the EALs are symptom-based. That is, the action level threshold is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because rt allows the full scope of variations in the types of events to be classified as emergencies. However, a purely symptom-based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no pre-determined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized. The CNS EAL methodology divides the EALs into 3 broad groups:

EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.
EALs applicable only under Modes 1, 2, or 3 - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode.
EALs applicable only under Modes 4, 5, or Defueled - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling, or Defueled mode.

The purpose of the groups is to avoid review of EALs that cannot be applicable in the current operating mode of the plant. This approach significantly minimizes the total number of EALs that must be reviewed by the Emergency Director for a given plant condition and thereby speeds identification of the appropriate applicable EAL. IEMERGENC Y PLAN REVISION 76 PAGE 1 04 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN ,,-..,_) Within each of the above three groups, assignment of EALs to categories/subcategories - Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. 5.7.1.1 EMERGENCY CLASSIFICATION PROCESS This procedure provides instruction for the evaluation, classification and declaration of an emergency at Cooper Nuclear Station. 5.7.2 EMERGENCY DIRECTOR EPIP This procedure provides a series of actions to be taken upon declaration of an Emergency Classification. Personnel shall be directed to use additional procedures to adequately respond to an emergency event. Certain actions may still need to be performed by the Shift Manager, as requested by the Emergency Director, after command and control of the emergency response has been transferred to the EOF. 5.7.3 INCIDENT COMMAND POST (ICP) HOSTILE ACTION BASED EVENT ROLES AND RESPONSIBILITIES This procedure provides guidance on actions to be taken after a Hostile Action Based (HAB) Event has occurred or is occurring, including Incident Command Post (ICP) and Cooper Nuclear Station (CNS) interface, communications with off-site agencies, and bringing off-site agencies into the Protected Area. 5.7.6 NOTIFICATION This procedure provides notification instructions to be followed upon declaration of an emergency condition. These include initial, follow-up, and event termination notifications to responsible State and Local Governmental Agencies and NRG notifications prior to TSC activation. Upon declaration of an emergency condition, all notifications and communications will be handled from the Control Room (CR) until the Technical Support Center (TSC) and/or the Emergency Operations Facility (EOF) are activated. All telephone numbers needed for notification or follow-up information transmission are in the Emergency Telephone Directory located in the Control Room, Technical Support Center, Emergency Operations Facility, and other designated areas. IEMERGENCY PLAN REVISION 76 PAGE 105 OF 192 I

APPENDI X A PROCEDURES IMPLEME NTING THE EMERGE NCY PLAN ') 5.7.7, ACTIVATI ON OF TSC 5.7.8 ACTIVATI ON OF OSC 5.7.8.1 ACTIVATI ON OF ALTERNATE OSC 5.7.8.2 ACTIVATI ON OF AL TERNATE OFF-SITE oscrrsc 5.7.9 ACTIVATI ON OF EOF These procedures describe the sequence of events and the staffing requirements for the activation of the TSC (5.7.7), OSC (5.7.8), Alternate OSC (5.7.8.1 ), Alternate Off-Site OSC!TSC (5.7.8.2) and the EOF (5.7.9). They provide further information on the functions of the CNS ERFs (see Section 7.2 of the Emergency Plan). 5.7.10 PERSONN EL ASSEMBL Y AND ACCOUN TABILITY This procedure describes the immediate and on-going emergenc y personnel assembly and accountability actions to be taken by all on-site personnel including ERO members, Security personnel, contractors, and visitors in the event of a station emergency. It also provides a means to ascertain the names of missing individuals within 30 minutes and account for all on-site individuals continuously thereafter. Since each site employee, Security Officer, visitor, and contractor is assigned a designate d assembly area and each area is assigned a Designated Assembly Area Supervisor (OMS) personnel accountability, in accordance with the discussions in Section 6.5 of the Emergency Plan, is assured. 5.7.11 EARLY DISMISSA UEVACUA TION OF SITE PERSONN EL As discussed in Section 6.5 of the Emergenc y Plan, in the event of an emergenc y situation it may be desirable to minimize the number of non-ERO personnel on-site. If the emergency involves a radiological release or the potential for a release, then evacuation of non-ERO personnel is desirable, or may be required to minimize exposure to radioactive material. This procedure provides an efficient means for evacuation of personne'I from isolated areas or from the plant site in its entirety. Furthermore, it provides a definition of the duties and responsibilities of designated supervisory personnel associated with site evacuation. 5.7.12 EMERGE NCY RADIATION EXPOSUR E CONTROL As indicated in Section 6.6 of the Emergenc y Plan, under certain emergenc y conditions it may become necessary for emergency workers to receive dose in excess of occupational limits established by 10CFR20. Dose limits for workers performing emergency services are contained in this procedure. As indicated in the Emergency Plan and Procedure, the Emergenc y Director has the authority to authorize dose in excess of occupational limits. This dose is only justifiable if it is determined that benefits are being achieved, the IEMERGENCY PLAN REVISION 76 PAGE 106 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN dose is commensurate with the significance of the objective, and every reasonable effort is being made to maintain emergency workers dose As Low As Reasonably Achievable (ALARA). The primary purpose of this procedure is to provide policy guidance, address required authorization, and set forth maximum criteria for emergency radiation dose control in the event emergency workers are required to exceed established annual dose limits. 5.7.13 PERSONNEL MONITORING AND DECONTAMINATION As discussed in Section 6.6 of the Emergency Plan, the objectives of personnel decontamination techniques are to promptly reduce radiation dose; to minimize the absorption of radionuclides, particularly radioiodine, into the body; and to prevent the spread of localized contamination. This procedure provides instructions for decontamination of station personnel during emergency conditions utilizing normal decontamination facilities or alternate areas if necessary. 5.7.14 STABLE IODINE THYROID BLOCKING (Kl) The purpose of this procedure is to define under what emergency conditions 0 Potassium Iodide (Kl) should be administered to station personnel and who has the authority to determine when and at what dosages Kl should be administered. The procedure also provides a discussion of the effectiveness of Kl, the recommended dosage, as well as any precautions and potential side effects. As indicated in Section 7.7 of the Emergency Plan, Kl tablets ar~stored in the CNS emergency lockers and are available on a voluntary basis to emergency response personnel as conditions dictate. These tablets will be dispensed only with the permission of the Emergency Director (also see Section 6.5.5 of the Emergency Plan). 5.7.15 OSC TEAM DISPATCH As defined in Section 6.5 of the Emergency Plan, if station personnel are unaccounted for in the initial or subsequent emergency accountability, the Emergency Director will assign an emergency team to locate, and if necessary, rescue them. This procedure provides guidance and requirements necessary to conduct efficient rescue and re-entry operations. It presents the organization and operation of Rescue and Re-Entry Teams and identifies the precautions which should be observed by the Rescue and Re-Entry Teams (including equipment carried during search and rescue operations). 5.7.16 RELEASE RATE DETERMINATION IEMERGENCY PLAN REVISION 76 PAGE 107 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN

~ This procedure describes methodology for the manual determination of airborne radioactive release rates from the Elevated Release Point, Reactor Building vent, the Turbine Building vent, and the Augmented.Radwaste Building vent utilizing effluent monitor readings.

Upon determination of release rates, actual or projected plume exposure dose may be calculated in accordance with EPIP 5.7.17, Dose Assessment or EPIP 5. 7.17.1, Dose Assessment (Manual). This dose provides a basis for relating plume exposure dose to the EPA Protective Action Guides (PAGs) in accordance with EPIP 5.7.20, Protective Action Recommendations (see Section 6.3 of the Emergency Plan). 5.7.17 DOSE ASSESSMENT This procedure provides a means for personnel to quickly predict off-site dose rates and integrated dose based on meteorological data, release rates, and dispersion. 5.7.17.1 DOSE ASSESSMENT (MANUAL) This procedure provides a means for personnel to perform a plume centerline or non-centerline dose assessment including dose projections from multiple source releases using hand calculation methods when the CNS-DOSE 0 computer program is not available. 5.7.18 OFF-SITE AND SITE BOUNDARY MONITORING In the event of an accidental radiological release, data obtained from off-site survey will be used to assess the magnitude of the release and to determine which off-site areas have been affected by the release. As indicated in Section 4.2 of the Emergency Plan, CNS will deploy Field Monitoring T earns for initial off-site monitoring prior to the arrival of responding state teams. The CNS Teams will remain in the field and assist the State(s) Teams as required. Data obtained through the off-site survey shall be utilized to determine actual release rates, deposition rates, and actual dose. Dose assessments provide a basis for decision making concerning recommendation of appropriate protective actions in accordance with EPIP 5.7.20, Protective Action Recommendations. Off-site survey data' will be used in conjunction with on-site release rate determination and dose assessment capabilities to accurately determine off-site consequences of an accident condition. This procedure describes the emergency off-site and site boundary radiological monitoring and field surveys to be undertaken in the event of an airborne release of radioactive gases from CNS. Instructions for the implementation of the program, locating sampling points, collecting samples, and performing field surveys are provided. 5.7.19 ON-SITE RADIOLOGICAL MONITORING IEMERGENCY PLAN REVISION 76 PAGE 108 OF 192 I

APPENDfXA PROCEDURES IMPLEMENTING THE EMERGENCY PLAN In the event of an accidental release involving radionuclides, data obtained from the on-site survey will be used to make initial assessments concerning the magnitude of the accident and decisions concerning evacuation of site personnel. Principal concerns for accidental radioactive releases, particularly gaseous releases, include limiting internal dose through appropriate respiratory protection equipment, anti-contamination clothing, limiting external dose by identifying areas of high radiation, and containment of the material to prevent the spreading of contamination or release to the environs. This procedure describes the emergency on-site radiological monitoring necessary to determine dose rates, airborne particulate, noble gas, and radioiodine activity levels due to an accidental release of radionuclides. The on-site survey entails the interior space of all station buildings. 5.7.20 PROTECTIVE ACTION RECOMMENDATIONS Dose estimates (which population groups may potentially receive) are calculated according to the dose assessment methodology described in EPIP 5.7.17, Dose Assessment or EPIP 5.7.17.1, Dose Assessment (Manual). These dose estimates are referred to as projected dose. A protective action is an action taken to avoid or reduce a projected dose when 0 the benefits derived from such action are sufficient to offset any undesirable features of the protective action. This procedure provides a basis for relating actual or projected plume exposure dose to the EPA Protective Action Guides (PAGs) in order to recommend the appropriate protective actions to the county or state governments (see Sections 6.5 and 6.6). 5.7.21 MAINTAINING EMERGENCY PREPAREDNESS - EMERGENCY EXERCISES, DRILLS, TESTS AND EVALUATIONS This procedure provides a means of ensuring the operational readiness and availability of equipment required for the immediate action steps of all four Emergency Classification action levels. This procedure also provides instructions for documenting the completion of periodic surveillances, tests, drills, and training to ensure availability, operability, and reliability. As an emergency situation progresses, conditions may arise which require augmentation of emergency equipment. The necessary equipment will be utilized on an as-needed basis to support the emergency operations (see Section 6.5.2 of the Emergency Plan). 5.7.23 ACTIVATION OF THE JIG This procedure describes the sequence of events and the staffing requirements for the activation of the Joint Information Center (JIG). It IEMERGENCY PLAN REVISION 76 PAGE 109 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN provides further infonnation on the functions of the JIG and its interaction with other CNS ERF's (see Section 7.2 of the Emergency Plan) and provides responsibilities of key emergency response personnel. 5.7.24 MEDICAL EMERGENCY As delineated in Sections 6.6.3 and 6.6.4 of the Emergency Plan, arrangements with local organizations have been made for the transportation and care of injured/contaminated personnel. This procedure details the treatment of injured personnel during a radiological emergency. The topics addressed by the procedure are: (1) aid to contaminated and non-contaminated injured personnel on-site, and (2) transportation and treatment of injured personnel. It also provides infonnation on local off-site facilities and the actions to be taken by off-site personnel. 5.7.25 RECOVERY OPERATIONS This procedure describes recovery operations necessary to identify the extent of station damage and radiological contamination (if any) and return the station to an operating status in compliance with the Technical Specifications. 0 Recovery operations will include measures taken during and immediately following the emergency, as well as, the longer term post-emergency efforts. These operations will be performed by station and other NPPD personnel, and if required, by contract technical and labor support. Manpower and equipment resources supporting the individual functional segments of the Recovery Organization will vary according to the severity of damage and specific situational needs. 5.7.26 LONG-TERM ENVIRONMENTAL MONITORING Methods to be used for evaluating long-term environmental consequences and analyses of trends involving key isotopes of radioactive material released from CNS are described in this procedure. Immediate collection and analysis of samples from impacted areas following a release shall be conducted in accordance with EPIP 5. 7 .18, Off-Site And Site Boundary Monitoring. Long-term environmental monitoring and trend analyses shall be conducted in accordance with EPIP 5.7.26. Appropriate protective measures are also discussed (also see Section 7.5.4 of the Emergency Plan). 5.7.27 ALERT AND NOTIFICATION SYSTEM 5.7.27.1 NOAA/EAS RADIO MALFUNCTION 5.7.27.2 FALSE ACTIVATION OF ALERT AND NOTIFICATION SYSTEM The purpose of these procedures are to describe the CNS Alert And IEMERGENCY PLAN REVISION 76 PAGE 110 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN Notification System within the Plume Exposure Pathway (10-Mile EPZ) of Cooper Nuclear Station. This system was set up to meet prompt notification requirements for Cooper Nuclear Station. These procedures also include response to inadvertent siren activation or EAS radio malfunctions. As indicated in Section 6.5 of the Emergency Plan, the CNS Early Warning System consists of fixed sirens covering areas of high population density and digitally-activated Emergency Alert System {EAS) radios in the low population density rural areas. The fixed siren system is composed of 21 pole-mounted sirens. Digital NOAA/EAS radio receivers are made available to residences located within the rural EPZ areas and outside the effective hearing range of the fixed sirens. These radios are pre-tuned to NOANEAS radio transmitter KWN41 (162.5 MHz) located at Shubert, NE. The receivers constantly monitor the broadcast frequency of the digital NOAA/EAS station, and activate upon receipt of the appropriate digital signals. 5.7.28 ADMINISTRATION OF POSITIONAL INSTRUCTION MANUALS (PIMS) This procedure provides guidance on the revision and control of the Positional Instructional Manuals {PIMs). This procedure shall ensure proposed changes 0 to the Pl Ms are properly evaluated and approved prior to implementation. This shall prevent any potential degradation to the CNS Emergency Plan and its associated Implementing Procedures. 5.7COMMUN COMMUNICATIONS As presented in Section 7.3 of the Emergency Plan, the Emergency Response Organization has available to it various types of communications equipment which allow for effective communication to both on-site and off-site groups. This procedure provides a description of these systems and very basic instructions for their use. 5.7ENS ENS COMMUNICATOR The procedure provides guidance to the ENS Communicator in the TSC for communicating with the NRC. 0-EP-01 EMERGENCY RESPONSE ORGANIZATION RESPONSIBILITIES This procedure defines the responsibilities of plant personnel in support of the Emergency Response Organization at CNS and also provides a description of the responsibilities of key emergency response personnel. 0-EP-02 CONFIGURATION CONTROL OF THE AUTOMATED NOTIFICATION SYSTEM (ANS) IEMERGENCY PLAN REVISION 76 PAGE 111 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN ~ This procedure provides the guidance needed to maintain the Automated Notification System. 0-EN-EP-306 DRILLS AND EXERCISES This procedure provides guidance for the planning, preparation, scheduling, conduct, evaluation, and documentation of ERO drills and exercises. 0-EN-EP-308 EMERGENCY PLANNING CRITIQUES This procedure describes the means for tracking and correcting deficiencies identified in ERO drills and exercises. TPP 101 EMERGENCY RESPONSE ORGANIZATION The Emergency Response Organization Training Program provides for initial qualification, requalification training and evaluation of CNS emergency response personnel. TPP 102 EMERGERNCY PREPAREDNESS STAFF TRAINING AND QUALIFICATION This training program procedure establishes the training and qualrfication requirements for the staff of the Emergency Preparedness (EP) Department 0 at Cooper Nuclear Station. EPDG 2 Att. H-1 CNS DRILL AND EXERCISE MANUAL - SCHEDULING This desk guide provides guidance for the development and maintenance of the drill and exercise schedule for Cooper Nuclear Station. EPDG 2 Att. H-2 CNS DRILL AND EXERCISE MANUAL - SCENARIO DEVELOPMENT This desk guide provides instruction for the development of scenarios used for integrated ERO drills or exercises. EPDG 2 Att. H-4 CNS DRILL AND EXERCISE MANUAL - DRILL AND EXERCISE PREPARATIONS This desk guide contains the checklists needed to prepare for the drills and exercises on the drill schedule. EPDG 2 Att. H-5 CNS DRILL AND EXERCISE MANUAL - CRITIQUE PROCESS This desk guide describes the process for the conduct and development of a drill or exercise critique. IEMERGENCY PLAN REVISION 76 PAGE 112 OF 192 I

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGE NCY PLAN ') EPIP/EMERGENCY PLAN CROSS-REFERENCE EPIP No. Emergency Plan Section 5.7.1 4.1, Tables 4.1-1 through 4.1-4 5.7.1.1 4.1, Tables 4.1-1 through 4.1-4 5.7.2 5.1.2, 6.2, 7.2 5.7.3 6.4, 6.5, 6.6 5.7.6 6.2, 7.4 5.7.7 5.2, 7.2.1 5.7.8 5.2, 7.2.2 5.7.8.1 5.2, 7.2.2 5.7.8.2 6.4, 6.5, 6.6 5.7.9 5.2, 7.2.3 5.7.10 6.5.3 5.7.11 6.5.4 5.7.12 6.5.5, 6.6.1, 6.6.2

J 5.7.13 5.7.14 6.5.5, 6.6.1, 6.6.2 6.5.5.1 5.7.15 6.5.1 5.7.16 4.2, 6.3, 7.5 5.7.17 4.2, 6.3, 7.5 5.7.17.1 4.2, 6.3, 7.5 5.7.18 4.2, 6.3, 7.5 5.7.19 4.2, 6.3, 7 .5 5.7.20 6.5, 6.6 5.7.21 6.5.2, 7.7, 7.8, Appendix E 5.7.23 5.3, 6.2.3, 7.2.4 5.7.24 6.6.2, 6.6.3, 6.6.4, 7.9 5.7.25 9.1, 9.2 5.7.26 7.5 5.7.27 6.5 5.7.27.1 6.5 5.7.27.2 6.5 IEMERGENCY PLAN REVISION 76 PAGE 113 OF 192 j

APPENDIX A PROCEDURES IMPLEMENTING THE EMERGENCY PLAN ') 5.7.28 None 5.7ENS 5.2.2.5 5.7COMMUN 7.3 0-EP-01 5.2 0-EP-02 6.2 0-EN-EP-306 8.2 O-EN-EP-308 8.2 TPP 101 8.1.1 TPP 102 8.1.2 EPDG 2 Att H-1 8.2 EPDG 2 Att H-2 8.2 EPDG 2 Att H-4 8.2 EPDG 2 Att H-5 8.2 C) IEMERGENCY PLAN REVISION 76 PAGE 114 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ) APPENDlX B CROSS INDEX TO NUREG--0664, REV 1 (FEMA REP 1) APPENDIX B Table of Contents Page Cross Reference NUREG-065 4, Rev. 1 (FEMA REP 1) to the Emergency Plan .......................................................................................................115 thru 157 Cross Reference 10CFR50.47. b to NUREG-0654/FEMA-REP-1, Rev. 1 .............................. 158 Cross Reference 10CFR50, Appendix E.IV to the Emergency Plan ....................................... 159 Cross Reference NEI 99-01, Rev 5 to EAL Number for Notification of Unusual Event Conditions ............................................................................................................................... 161 Cross Reference NEI 99-01, Rev 5 to EAL Number for Alert Conditions ............................... 163 Cross Reference NEI 99-01, Rev 5 to EAL Number for Site Area Emergency Conditions ............................................................................................................164 Cross Reference NEI 99-01, Rev 5 to EAL Number General Emergency Conditions ............ 165 ,:) Cross Reference NEI 10-05, Rev 0 to Assessment of On-Shift Emergency Response Organization Staffing and Capabilities .................................................................... .47 IEMERGENCY PLAN REVISION 76 PAGE 115 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ~ PLANNING STANDARDS AND EVALUATION CRITERIA A. ASSIGNMENT OF RESPONSIBILITY (ORGANIZATION CONTROL) PLANNING STANDARD Primary responsibilities for emergency response by the nuclear facility licensee, and by State and Local Organizations within the Emergency Planning Zones have been assigned, the emergency responsibilities of the various supporting organizations have been specifically established, and each principal response organization has staff to respond and to augment its initial response on a continuous basis. CROSS REFERENCE EVALUATION CRITERIA TO PLAN 1.

a. Each plan shall identify the State, local, Federal Sections 5, 5.1, 5.2, 5.3, 5.4, and private sector organizations (including utilities), 7.9.1 that are intended to be part of the overall response Appendices D, F organization for Emergency Planning Zones (see Appendix 5).

0 b. Each organization and sub-organization having an operational role shall specify its concept of Section 5 operations, and its relationship to the total effort. C. Each plan shall illustrate these interrelationships in Figures 5.2-1 through 5.2-4, a block diagram. 5.3-1, 5.4-1, 5.4-2

d. Each organization shall identify a specific individual Sections 5.1.2, 5.2 by title who shall be in charge of the emergency Figures: 5.2-1, 5.2-2, 5.2-3, response. 5.2-4

e. Each organization shall provide for 24 hour per day Sections 5.1, 5.1.2.1 , 5.1.2.6, emergency response, including 24 hour per day 5.2.2.5, 5.2.3.3 manning of communications links.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN 2.

a. Each organization shall specify the functions and Not Applicable to Licensee responsibilities for major elements and key individuals by title, of emergency response, including the following: Command and Control, Alerting and Notification, Communications, Public Information, Accident Assessment, Public Health and Sanitation, Social Services, Fire and Rescue, Traffic Control, Emergency Medlcal Services, Law Enforcement, Transportation, Protection Response (including authority to request Federal assistance and to initiate other protective actions), and Radiological Exposure Control. The description of these functions shall include a clear and concise summary such as a table of primary and support responsibilities using the agency as one axis, and the function as the other (see Section B for licensee).

b. Each plan shall contain (by reference to specific Not Applicable to Licensee acts, codes or statutes) the legal basis for such authorities.

3. Each plan shall include written agreements referring to Appendix D the concept of operations developed between Federal, State, and local agencies and other support organizations having an emergency response role within the Emergency Planning Zones. The agreements shall identify the emergency measures to be provided and the mutually acceptable criteria for their implementation, and specify the arrangements for exchange of information. These agreements may be provided in an appendix to the plan or the plan itself may contain descriptions of these matters and a signature page in the plan may serve to verify the agreements. The signature page format is appropriate for organizations where response functions are covered by laws, regulations or executive orders where separate written agreements are not necessary.

IEMERGENCY PLAN REVISION 76 PAGE 117 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

4. Each principal organization shall be capable of Section 5.2.3.1 O continuous (24-hour) operations for a protracted period.

The individual in the principal organization who will be responsible for assuring continurty of resources (technical, administrative, and material) shall be specmed by title. IEMERGENCY PLAN REVISION 76 PAGE 11 8 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) B. ON-SITE EMERGENCY ORGANIZATION PLANNING STANDARD On-shift facility licensee responsibilities for emergency response are unambiguously defined, adequate staffing to provide initial facility accident response in key functional areas is maintained at all times, timely augmentation of response capabilities is available, and the interfaces among various on-site response activities and off-site support and response activities are specified. CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each licensee shall specify the on-site emergency Sections 5, 5.1 organization of plant staff personnel for all shifts and Figure 5.2-1 its relation to the responsibilities and duties of the normal staff complement.

2. Each licensee shall designate an individual as Section 5.1 Emergency Coordinator who shall be on shift at all Figure 5.2-1 times and who shall have the authority and responsibility to immediately and unilaterally initiate any emergency actions, including providing protective action recommendations to authorities responsible for C) implementing off-site emergency measures.

3. Each licensee shall identify a line of succession for the Sections 5.1.1.2, 5.1.2.1 Emergency Coordinator position and identify the specific conditions for higher level utility officials assuming this function.

4. Each licensee shall establish the functional Sections 5.1.2, 5.2, 5.2.1 responsibilities assigned to the Emergency Coordinator and shall clearly specify which responsibilities may not be delegated to other elements of the emergency organization. Among the responsibilities which may not be delegated shall be the decision to notify and to recommend protective actions to authorities responsible for off-site emergency measures.

IEMERGENCY PLAN REVISION 76 PAGE 119 OF 192 j

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

5. Each licensee shall specify the positions or title and Section 5.2 major tasks to be performed by the persons to be Figures 5.2-1, 5.2-2, 5.2-3, assigned to the functional areas of emergency activity. 5.2-4, 5.3-1, 5.4-1 For emergency situations, specific assignments shall be made for all shifts and for plant staff members, both on-site and away from the site. These assignments shall cover the emergency functions in Table 8-1 entitled, "Minimum Staffing Requirements for Nuclear Power Plant Emergencies". The minimum on-shift staffing levels shall be as indicated in Table 8-1. The licensee must be able to augment on-shift capabilities within a short period after declaration of an emergency.

This capability shall be as indicated in Table 8-1. The implementation schedule for Licensed Operators, Auxiliary Operators and the Shift Technical Advisor on shift shall be as specified in the July 31, 1980, letter to all power reactor licensees. Any deficiencies in the other staff requirements of Table 8-1 must be capable of augmentation within 30 minutes by September 1, 1981, and such deficiencies must be fully removed by July 1, 1982.

6. Each licensee shall specify the interfaces between and Figures 5.4-1, 5.4-2 among the on-site functional areas of emergency activity, licensee headquarters support, local services support, and State and Local Government Response Organization. This shall be illustrated on a block diagram and shall include the on-site Technical Support Center and the Operational Support (assembly) Center and the licensee's near site Emergency Operations Facility (EOF).

7. Each licensee shall specify the Corporate Management, Section 5.3 Administrative and Technical Support personnel who Figure 5.3-1 will augment the plant staff as specified in the table entitled "Minimum Staffing Requirements for Nuclear Power Plant Emergencies", (Table B-1) and in the following areas:

a. Logistics support for emergency personnel (e.g., Section 5.2.3.10 transportation, communications, temporary quarters, food and water, sanitary facilities in the field, and special equipment and supplies procurement);

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

b. Technical support for planning and re-entry/recovery Section 9 operations;

c. Management level interface with governmental Sections 5.2, 5.2.3 authorities; and

d. Release of information to news media during an Sections 5.3.1, 5.3.2 emergency (coordinated with governmental authorities).

8. Each licensee shall specify the contractor and private Section 5.3.3 organizations who may be requested to provide Appendix D technical assistance to and augmentation of the emergency organization.

9. Each licensee shall identify the services to be provided Sections 5.4, 6.6.3, 7.9.1 by local agencies for handling emergencies (e.g., police, Appendix D, F ambulance, medical, hospital, and fire-fighting organizations shall be specified). The licensee shall provide for transportation and treatment of injured personnel who may also be contaminated. Copies of the arrangements and agreements reached with contractor, private, and local support agencies shall be appended to the plan. The agreements shall delineate the authorities, responsibilities, and limits on the actions of the contractor, private organization, and local services support groups.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1)

,,,-,,I C. EMERGENCY RESPONSE SUPPORT AND RESOURCES PLANNING STANDARD Arrangements for requesting and effectively using assistance resources have been made, arrangements to accommodate State and Local Staff at the licensee's near-site Emergency Operations Facility have been made, and other organizations capable of augmenting the planned response have been identified.

CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. The Federal government maintains in-depth capability to assist licensees, States and local governments through the Federal Radiological Monitoring and Assessment Plan (formerly Radiological Assistance Plan (RAP) and lnteragency Radiological Assistance Plan (IRAP). Each State and licensee shall make provisions for incorporating the Federal response capability into its operation plan, including the following:

a. Specific persons by title authorized to request Sections 5.2.3, 7.2.3 Federal assistance; see A.1.d., A.2.a.

'.:)

b. Specific Federal resources expected, including Figures 5.4-1, 5.4-2 expected times of arrival at specific nuclear facility sites; and C. Specific licensee, State and Local resources Section 7 available to support the Federal response (e.g., air fields, command posts, telephone lines, radio frequencies and telecommunications centers).

2.

a. Each principal off-site organization may dispatch Not Applicable to Licensee representatives to the licensee's near-site Emergency Operations Facility (State technical Analysis representatives at the near-site EOF are preferred).

b. The licensee shall prepare for the dispatch of a NIA representative to principal Off-Site Governmental Emergency Operations Centers.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

3. Each organization shall identify radiological laboratories Sections 5.3.3.3, 7.5.7 and their general capabilities and expected availability to provide radiological monitoring and analyses services which can be used in an emergency.

4. Each organization shall identify nuclear and other Section 5.3.3 facilities, orgarJizations or individuals which can be Appendix D relied upon in an emergency to provide assistance.

Such assistance shall be identified and supported by appropriate letters of agreement. IEMERGENCY PLAN REVISION 76 PAGE 123 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1)

~  D. EMERGENCY CLASSIFICATION SYSTEM PLANNING STANDARD A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and Local Response Plans call for reliance on information provided by facility licensees for determinations of minimum initial off-site response measures.

CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. An emergency classification and emergency action level Section 4 scheme as set forth in Appendix 1 must be established by the licensee. The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant Emergency Procedures. The plan shall identify the parameter values and equipment status for each emergency class.

2. The initiating conditions shall include the example Tables 4.1-1, 4.1-2, 4.1-3, conditions found in Appendix 1 and all postulated 4.1-4 accidents in the Final Safety Analysis Report (FSAR)

Q for the nuclear facility.

3. Each State and Local Organization shall establish an Not Applicable to Licensee emergency classification and emergency action level scheme consistent with that established by the facility licensee.

4. Each State and Local Organization should have Not Applicable to Licensee procedures in place that provide for emergency actions to be taken which are consistent with the emergency actions recommended by the nuclear facility licensee, taking into account local off-site conditions that exist at the time of the emergency.

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APPENDIX 8 CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ~ E. NOTIFICATION METHODS AND PROCEDURES PLANNING STANDARDS Procedures have been established for notification, by the licensee of State and Local Response Organizations and for notification of emergency personnel by all response organizations; the content of initial and follow-up messages to response organizations and the public has been established; and means to provide early notification and clear instruction to the populace within the plume exposure pathway Emergency Planning Zone. CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each organization shall establish procedures which Sections 6.2.4, 7.4 describe mutually agreeable bases for notification of Figure 7.4-1 response organizations consistent with the emergency classification and action level scheme set forth in Appendix 1. These procedures shall include means for verification. Specific details need not be included in the plan.

2. Each organization shall establish procedures for Section 6.1, 6.2, 7.4 alerting, notifying, and mobilizing emergency response Figure 7.4-1 C) personnel.

3. The licensee in conjunction with State and Local Section 6.2.4 Organizations shall establish the contents of the initial emergency messages to be sent from the plant. These measures shall contain information about the class of emergency, whether a release is taking place, potentially affected population and areas, and whether protective measures may be necessary.

4. Each licensee shall make provisions for follow-up Section 6.2.4 messages from the facility to off-site authorities which shall contain the following information if it is known and appropriate:

a. Location of incident and name and telephone Section 6.2.4 number (or communications channel identification) of caller;

b. Date/time of incident; Section 6.2.4

c. Class of emergency; Section 6.2.4 IEMERGENCY PLAN REVISION 76 PAGE 125 OF 192 j

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1)

~                       EVALUATION CRITERIA CROSS REFERENCE TO PLAN

d. Type of actual or projected release (airborne, Section 6.2.4 waterborne, surface spill), and estimated duration/impact times;

e. Estimate of quantity of radioactive material released Section 6.2.4 or being released and the points and height of releases;

f. Chemical and physical form of released Section 6.2.4 material, including estimates of the relative quantities and concentration of noble gases, iodines, and particulates;

g. Meteorological conditions at appropriate levels (wind Section 6.2.4 speed, direction (to and from), indicator of stability, precipitation, if any);

h. Actual or projected dose rates at site boundary; Section 6.2.4 projected integrated dose at site boundary;

.'.~

i. Projected dose rates and integrated dose at the Section 6.2.4 projected peak and at 2, 5, and 10 miles, including sector(s) affected;

j. Estimate of any surface radioactive Section 6.2.4 contamination in-plant, on-site or off-site;

k. Licensee emergency response actions underway; Section 6.2.4 I. Recommended emergency actions, including Section 6.2.4 protective measures;

m. Request for any needed on-site support by off-site Section 6.2.4 organizations; and

n. Prognosis for worsening or termination of event Section 6.2.4 based on plant information.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

5. State and Local Government Organizations shall Not Applicable to Licensee establish a system for disseminating to the public appropriate information contained in initial and follow-up messages received from the licensee including the appropriate notification to appropriate broadcast media (e.g., the Emergency Broadcast System (EBS)).

6. Each organization shall establish administrative and Section 6.5 physical means, and the time required for notifying and providing prompt instructions to the public within the plume exposure pathway Emergency Planning Zone (see Appendix 3.). It shall be the licensee's responsibility to demonstrate that such means exist, regardless of who implements this requirement. It shall be the responsibility of the State and Local Governments to activate such a system.

7. Each organization shall provide written messages Section 6.5 intended for the public, consistent with the licensee's classification scheme. In particular, draft messages to the public giving instructions with regard to specific protective actions to be taken by occupants of affected areas shall be prepared and included as part of the State and Local Plans. Such messages should include the appropriate aspects of sheltering, ad hoc respiratory protection (e.g., handkerchief over mouth, thyroid blocking, or evacuation). The role of the licensee is to provide supporting information for the messages. For ad hoc respiratory protection see "Respiratory Protective Devices Manual" American Industrial Hygiene Association, 1963 pp. 123-126.

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") F. EMERGENCY COMMUNICATIONS PLANNING STANDARD Provisions exist for prompt communications among principal response organizations to emergency personnel and to the public.

CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. The communication plans for emergencies shall include organizational titles and alternates for both ends of the communication links. Each organization shall establish reliable primary and backup means of communication for licensees, local, and State response organizations.

Such systems should be selected to be compatible with one another. Each plan shall include:

a. Provision for 24-hour per day notification to and Sections 5.1, 5.1.2.1, 5.1.2.6, activation of the State/local emergency response 5.2.2.5, 5.2.3.3, 7.3.

network; and at a minimum, a telephone link and Table 7.1-1 alternate, including 24-hour per day manning of communications links that initiate emergency response actions;

b. Provision for communications with contiguous Sections 6.2.4, 7.1, 7.2, 7.3 State/local governments within the Emergency Planning Zones;

c. Provision for communications as needed with Sections 6.2.5, 7.1, 7.2, 7.3 Federal emergency response organizations;

d. Provision for communications between the nuclear Table 7.1-1 facility and the licensee's near-site Emergency Operations Facility, State and Local Emergency Operations Centers, and Radiological Monitoring Teams;

e. Provision for alerting or activating emergency Sections 6.2, 6.2.1, 6.2.2, personnel in each response organization; and 6.2.3, 6.2.4, 6.2.5 IEMERGENCY PLAN REVISION 76 PAGE 128 OF 192 I

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f. Provision for communication by the licensee Sections 6.2.5, 7 .1, 7.2, 7 .3 with NRC headquarters and NRC Regional Office Table 7.1-1 Emergency Operations Centers and the licensee's near-site Emergency Operations Facility and Radiological Monitoring Team assembly area.

2. Each organization shall ensure that a coordinated Sections 6.6.3, 6.6.4 communication link for fixed and mobile medical support facilities exists.

3. Each organization shall conduct periodic testing of the Section 8.2.3.5 entire Emergency Communications System (see Evaluation Criteria H.10, 2.a and Appendix 3).

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ~, G. PUBLIC EDUCATION AND INFORMATION PLANNING STANDARD lnfonnation is made available to the public on a periodic basis on how they will be notified and what their initial actions should be in an emergency (e.g., listening to a local broadcast station and remaining indoors), the principal points of contact with the news media for dissemination of infonnation during an emergency (including the physical location or locations) are established in advance, and procedures for coordinated dissemination of infonnation to the public are established. CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each organization shall provide a coordinated periodic Section 8.1.4 (at least annually) dissemination of infonnation to the public regarding how they will be notified and what their actions should be in an emergency. This infonnation shall include, but not necessarily be limited to:

a. Educational infonnation on radiation;

b. Contact for additional infonnation; C)

c. Protective measures (e.g., evacuation routes and relocation centers, sheltering, respiratory protection, radio-protective drugs); and

d. Special needs of the handicapped.

Means for accomplishing this dissemination may include, but are not necessarily limited to: infonnation in the telephone book; periodic infonnation in utility bills; posting in public areas; and publications distributed on an annual basis. IEMERGENCY PLAN REVISION 76 PAGE 130 OF 192 I

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~                         EVALUATION CRITERIA CROSS REFERENCE TO PLAN

2. The public information program shall provide the Section 8.1.4 pennanent and transient adult population within the plume exposure EPZ an adequate opportunity to become aware of the infonnation annually. The programs should include provision for written material that is likely to be available in a residence during an emergency. Updated infonnation shall be disseminated at least annually. Signs or other measures (e.g.,

decals, posted notices or other means, placed in hotels, motels, gasoline stations and phone booths) shall also be used to disseminate to any transient population within the plume exposure pathway EPZ appropriate infonnation that would be helpful if an emergency or accident occurs. Such notices should refer the transient to the telephone directory or other source of local emergency information and guide the visitor to appropriate radio and television frequencies. 3. /'.:) a. Each principal organization shall designate the Sections 5.3.1, 7.2.4, 8.1.5 points of contact and physical locations for use by Figure 7.2-4 news media during an emergency.

b. Each licensee shall provide space which may be Section 7 .2.4 used for a limited number of the news media at the Figure 7.2-4 near-site Emergency Operations Facility.

4.

a. Each principal organization shall designate a Sections 5.3.2, 7.2.4 spokesperson who should have access to all necessary information.

b. Each organization shall establish arrangements for Sections 5.3.2, 7.2.4 timely exchange of information among designated _/

spokespersons. C. Each organization shall establish coordinated Section 5.3.1, 5.3.1.3, 7 .2.4 arrangements for dealing with rumors. IEMERGENCY PLAN REVISION 76 PAGE 131 OF 192 I

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5. Each organization shall conduct coordinated programs Section 8.1.5 at least annually to acquaint news media with the emergency plans, information concerning radiation, and points of contact for release of public information in an emergency.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1)

-----) H EMERGENCY FACILITIES AND EQUIPMENT PLANNING STANDARD Adequate emergency facilities and equipment to support the emergency response are provided and maintained.

CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each licensee shall establish a Technical Support Sections 7.2.1, 7.2.2 Center and an On-Site Operations Support Center Figures 7.2-1, 7.2-2 (assembly area) in accordance with NUREG-0696, Revision 1.

2. Each licensee shall establish an Emergency Operations Section 7.2.3 Facility from which evaluation and coordination of all Figures 7 .2-3 licensee activities related to an emergency is to be carried out and from which the licensee shall provide information to Federal, State, and local authorities responding to radiological emergencies in accordance with NUREG-0696, Revision 1.

3. Each organization shall establish an Emergency Not Applicable to Licensee Operations Center for use in directing and controlling response functions.

4. Each organization shall provide for timely activation and Section 6.2 staffing of the facilities and centers described in the plan.

5. Each licensee shall identify and establish on-site Section 7 .5 monitoring systems that are to be used to initiate emergency measures in accordance with Appendix 1, as well as those to be used for conducting assessment.

The equipment shall include:

a. Geophysical phenomena monitors (e.g., Sections 7.5.1, 7.5.2, 7.5.3 meteorological, hydrologic, seismic);

b. Radiological monitors (e.g., process, area, Section 7.5.4, 7.5.5 emergency, effluent, wound and portable monitors and sampling equipment);

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c. Process monitors (e.g., Reactor Coolant System Section 7.5.5.6 pressure and temperature, containment pressure -

and temperature, liquid levels, flow rates, status or lineup of equipment components); and

d. Fire and combustion products detectors. Section 7.6

6. Each licensee shall make provision to acquire data from or for emergency access to off-site monitoring and analysis equipment including:

a. Geophysical phenomena monitors (e.g., Sections 7.5, 7.5.1, 7.5.2, meteorological, hydrologic, seismic); 7.5.3

b. Radiological monitors including ratemeters and Section 7.5.4 sampling devices. Dosimetry shall be provided and shall meet, as a minimum, the NRC Radiological Assessment Branch Technical Position for the Environmental Radiological Monitoring Program; and

c. Laboratory facilities, fixed or mobile. Section 7.5.7

7. Each organization, where appropriate, shall provide for Section 7.5.6 off-site radiological monitoring equipment in the vicinity of the nuclear facility.

8. Each licensee shall provide meteorological Section 7.5.2 instrumentation and procedures which satisfy the criteria in Appendix 2, and provisions to obtain representative current meteorological information from other sources.

9. Each licensee shall provide for an On-Site Operations Section 7.2.2 Support Center (assembly area) which shall have adequate capability, and supplies, including, for example, respiratory protection, protective clothing, portable lighting, portable radiation monitory equipment, cameras, and communications equipment for personnel present in the assembly area.

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10. Each organization shall make provisions to inspect, Sections 7.7, 8.6 inventory and operationally check emergency Appendix E equipment/instruments at least once each calendar quarter and after each use. There shall be sufficient reserves of instruments/equipment to replace those which are removed from emergency kits for calibration or repair. Calibration of equipment shall be at intervals recommended by the supplier of the equipment.

11. Each plan shall, in an appendix, include identification of Table 7.1-1 emergency kits by general category (protective Appendix E equipment, communications equipment, radiological monitoring equipment, and emergency supplies).

12. Each organization shall establish a central point Section 7.2.3 (preferably associated With the licensee's near-site Emergency Operations Facility), for the receipt and analysis of all field monitoring data and coordination of sample media.

0 IEMERGENCY PLAN REVISION 76 PAGE 135 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ) I. ACCIDENT ASSESSMENT PLANNING STANDARD Adequate methods, systems, and equipment for assessing and monitoring actual or potential off-site consequences of a radiological emergency condition are in use. CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each licensee shall identify plant system and effluent Tables 4.1-1, 4.1-2, 4.1-3, parameter values characteristic of a spectrum of 4.1-4 off-normal conditions and accidents, and shall identify the plant parameter values or other information which correspond to the example initiating conditions of Appendix 1. Such parameter values and the corresponding emergency class shall be included in the appropriate Facility Emergency Procedures. Facility Emergency Procedures shall specify the kinds of instruments being used and their capabilities.

2. On-site capability and resources to provide initial values Sections 4.2, 6.3, 7.5 and continuing assessment throughout the course of an accident shall include post-accident sampling capability, radiation and effluent monitors, in-plant iodine instrumentation, and containment radiation monitoring in accordance with NUREG-0578, as elaborated in the NRG letter to all power reactor licensees dated October 30, 1979.

3. Each licensee shall establish methods and techniques to be used for determining:

a. The source term of releases of radioactive material Sections 6.3.3, 7.5 within plant systems. An example is the relationship between the containment radiation monitor(s) reading(s) and radioactive material available for release from containment.

b. The magnitude of the release of radioactive Sections 6.3.3, 7.5 materials based on plant system parameters and effluent monitors.

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4. Each licensee shall establish the relationship between Sections 6.3.3, 7 .5 effluent monitor readings and on-site and off-site exposures and contamination for various meteorological conditions.

5. Each licensee shall have the capability of acquiring and Sections 6.3.2, 7.5 evaluating meteorological information sufficient to meet the criteria of Appendix 2. There shall be provisions for access to meteorological information by at least the near-site Emergency Operations Facility, the Technical Support Center, the Control Room, and an off-site NRC Center. The licensee shall make available to the State suitable meteorological data processing interconnections which will permit independent analysis by the State, of facility-generated data in those States with the resources to effectively use this information.

6. Each licensee shall establish the methodology for Sections 4.2, 6.3.3 determining the release rate/projected doses if the instrumentation used for assessment are off-scale or inoperable.

7. Each organization shall describe the capability and Sections 4.2, 6.3 resources for field monitoring within the plume exposure Emergency Planning Zone which are an intrinsic part of the concept of operations for the facility.

8. Each organization, where appropriate, shall provide Sections 4.2, 6.3, 7 .5 methods, equipment, and expertise to make rapid Table 6.3-1 assessments of the actual or potential magnitude and locations of any radiological hazards through liquid or gaseous release pathways. This shall include activation, notification means, field team composition, transportation, communication, monitoring equipment and estimated deployment times.

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9. Each organization shall have a capability to detect and Sections 6.3.3, 7.5.5.5 measure radioiodine concentrations in air in the plume exposure EPZ as low as 10-7 µCi/cc (microcuries per cubic centimeter) under field conditions. Interference from the presence of noble gas and background radiation shall not decrease the stated minimum detectable activity.

10. Each organization shall establish means for relating the Sections 6.3.3, 6.5 various measured parameters (e.g., contamination Tables 6.3-1, 6.4-1 levels, water and air activity levels) to dose rates for key isotopes (i.e., those given in Table 3, Page 18) and gross radioactivity measurements. Provisions shall be made for estimating integrated dose from the projected and actual dose rates and for comparing these estimates with the protective action guides. The detailed provisions shall be described in separate procedures.

11. Arrangements to locate and track the airborne Not Applicable to Licensee radioactive plume shall be made, using either or both Federal and State resources.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) r') J. PROTECTIVE RESPONSE PLANNING STANDARD A range of protective actions have been developed for the plume exposure pathway EPZ for emergency workers and the public. Guidelines for the choice of protective actions during an emergency, consistent with Federal guidance, are developed and in place, and protective actions for the ingestion exposure pathway EPZ appropriate to the locale have been developed. CROSS REFERENCE EVALUATION CRITERIA TO PLAN r 1. Each licensee shall establish the means and time Sections 6.1, 6.5 required to warn or advise on-site individuals and individuals who may be in areas controlled by the Operator, including:

a. Employees not having emergency assignments;

b. Visitors; Q c. Contractor and construction personnel; and

d. Other persons who may be in the public access areas on or passing through the site or within the owner-controlled area.

2. Each licensee shall make provisions for evacuation Section 6.5.4 routes and transportation for on-site individuals to some suitable off-site location, including alternatives for inclement weather, high traffic density and specific radiological conditions.

3. Each licensee shall provide for radiological monitoring Section 6.6.2 of people evacuated from the site.

4. Each licensee shall provide for the evacuation of on-site Section 6.5.4 non-essential personnel in the event of a Site or General Emergency and shall provide a decontamination capability at or near the monitoring point specified in J.3.

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5. Each licensee shall provide for a capability to account Section 6.5.3 for all individuals on-site at the time of the emergency and ascertain the names of missing individuals within 30 minutes of the start of an emergency and account for all on-site individuals continuously thereafter.

6. Each licensee shall, for individuals remaining or arriving Sections 6.5.2, 6.5.5.1 on-site during the emergency, make provisions for:

a. Individual respiratory protection;

b. Use of protective clothing; and

c. Use of radioprotective drugs (e.g., individual thyroid protection).

7. Each licensee shall establish a mechanism for Section 6.5 recommending protective actions to the appropriate Table 6.4-1 State and Local Authorities. These shall include Emergency Action Levels corresponding to projected dose to the population-at-risk, in accordance with Appendix 1 and with the recommendation s set forth in Tables 2.1 and 2.2 of the Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (EPA400-R-92-001 ). As specified in Appendix 1, prompt notification shall be made directly to the off-site authorities responsible for implementing protective measures within the plume exposure pathway Emergency Planning Zone.

8. Each licensee's plan shall contain time estimates for Appendix C evacuation within the plume exposure EPZ. These shall be in accordance with Appendix 4.

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9. Each State and Local Organization shall establish a Not Applicable to Licensee capability for implementing protective measures based upon protective action guides and other criteria. This shall be consistent with the recommendations of EPA regarding exposure resulting from passage of radioactive airborne plumes (EPA-400-R-92-001) and with those of DHEW (DHHS)/FDA regarding radioactive contamination of human food and animal feeds as published in the Federal Register of December 15, 1978 (43 FR 58790).

10. The organization's plans to implement protective measures for the plume exposure pathway shall include:

a. Maps showing evacuation routes, evacuation areas, Appendix C preselected radiological sampling and monitoring points, relocation centers in host areas, and shelter areas (identification of radiological sampling and monitoring points shall include the designators in Table J-1 or an equivalent uniform system described in the plan);

b. Maps showing population distribution around the Appendix C nuclear facility. This shall be by evacuation areas (licensees shall also present the information in a sector format);

c. Means for notifying all segments of the transient and Section 6.5 resident population;

d. Means for protecting those persons whose mobility Not Applicable to Licensee may be impaired due to such factors as institutional or other confinement;

e. Provisions for the use of radio-protective drugs, Not Applicable to Licensee particularly for emergency workers and institutionalized persons within the plume exposure EPZ whose immediate evacuation may be infeasible or very difficult, including quantities, storage, and means of distribution. '

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f. State and Local Organizations' Plans should Not Applicable to Licensee include the method by which decisions by the State Health Department for administering radio-protective drugs to the general population are made during an emergency and the pre-detennined conditions under which such drugs may be used by off-site emergency workers; 1 1 See DHEW (new DHHS) Federal Register notice of December 15, 1978 (43 FR 58798) entitled "Potassium Iodide as a Thyroid Blocking Agent in a Radiation Emergency." Other guidance concerning the storage, stockpiling, and conditions for use of this drug by the general public, is now under development by the Bureau of Drugs, DHHS.

g. Means of relocation; Not Applicable to Licensee

h. Relocation centers in host areas which are at least Not Applicable to Licensee 5 miles and preferably 10 miles, beyond the boundaries of the plume exposure emergency planning zone (see K.8);

i. Projected traffic capacities of evacuation routes Not Applicable to Licensee under emergency conditions;

j. Control of access to evacuated areas and Not Applicable to Licensee organization responsibilities for such control;

k. Identification of and means for dealing with potential Not Applicable to Licensee impediments (e.g., seasonal impassability of roads) to use of evacuation routes and contingency measures; I. Time estimates for evacuation of various Not Applicable to Licensee sectors and distances based on a dynamic analysis (time-motion study under various conditions) for the plume exposure pathway emergency planning zone (see Appendix 4); and I EMERGENCY PLAN REVISION 76 PAGE 142 OF 192 I

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m. The bases for the choice of recommended protective Section 6.5 actions from the plume exposure pathway during Tables 6.4-1, 6.4-2 emergency conditions. This shall include expected local protection afforded 2 in residential units or other shelter for direct and inhalation exposure, as well as evacuation time estimates.

2 The following reports may be considered in determining protection afforded. (1l "Public Protection Strategies for Potential Nuclear Reactor Accidents Sheltering Concepts with Existing Public and Private Structures" (SAND 771725), Sandia Laboratory.

           <2> "Examination of Off-Site Radiological Emergency Measures for Nuclear Reactor Accidents Involving Core Melt" (SAND 780454), Sandia Laboratory.

(3) "Protective Action Evaluation Part 11, Evacuation and Sheltering as Protective Actions Against Nuclear Accidents Involving Gaseous Releases" (EPA 520/1780018 U.S. Environmental Protection Agency.

11. Each State shall specify the protective measures to be Not Applicable to Licensee used for the ingestion pathway, including the methods for protecting the public from consumption of contaminated foodstuffs. This shall include criteria for

() deciding whether dairy animals should be put on stored feed. The plan shall identify procedures for detecting contamination, for estimating the dose commitment consequences of uncontrolled ingestion, and for imposing protection procedures such as impoundment, decontamination, processing, decay, product diversion, and preservation. Maps for recording survey and monitoring key land use data (e.g., farming), dairies, food processing plants, water sheds, water supply intake, and treatment plants and reservoirs shall be maintained. Provisions for maps showing detailed crop infonnation may be by including reference to their availability and location and a plan for their use. The maps shall start at the facility and include all of the 50-mile ingestion pathway EPZ. Up-to-date lists of the name and location of all facilities which regularly process milk products and other large amounts of food or agricultural products originating in the ingestion pathway Emergency Planning Zone, but located elsewhere, shall be maintained. IEMERGENCY PLAN REVISION 76 PAGE 143 OF 192 I

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12. Each organization shall describe the means for Not Applicable to Licensee registering and monitoring of evacuees at relocation centers in host areas. The personnel and equipment available should be capable of monitoring within about a 12-hour period all residents and transients in the plume exposure EPZ arriving at relocation centers.

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APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ') K. RADIOLOGICAL EXPOSURE CONTROL PLANNING STANDARD Means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides. CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each licensee shall establish on-site exposure Section 6.6.1 guidelines consistent with EPA Emergency Worker Table 6.4-1 and Lifesaving Activity Protective Actions Guides (EPA 400-R-92-001) for:

a. Removal of injured persons;

b. Undertaking corrective actions; C. Performing assessment actions;

(-) d. Providing first aid;

e. Performing personnel decontamination;

f. Providing ambulance service; and

g. Providing medical treatment services.

2. Each licensee shall provide an on-site radiation Section 6.6.1 protection program to be implemented during emergencies, including methods to implement exposure guidelines. The plan shall identify individual(s), by position or title, who can authorize emergency workers to receive doses in excess of 10CFR Part 20 limits.

Procedures shall be worked out in advance for permitting on-site volunteers to receive radiation exposures in the course of carrying out lifesaving and other emergency activities. These procedures shall include expeditious decision making and a reasonable consideration of relative risks. I EMERGENCY PLAN REVISION 76 PAGE 145 OF 192 I

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a. Each organization shall make provision for Section 6.6.1 24-hour-per-day capability to determine the doses received by emergency personnel involved in any nuclear accident, including volunteers. Each organization shall make provisions for distribution of dosimeters, both self-reading and permanent record devices.

b. Each organization shall ensure that dosimeters are Section 6.6.1 read at appropriate frequencies and provide for maintaining dose records for emergency workers involved in any nuclear accident.

4. Each State and Local Organization shall establish the Not Applicable to Licensee decision chain for authorizing emergency workers to incur exposures in excess of the EPA General Public Protective Action Guides (i.e., EPA PAGs for 0 emergency workers and lifesaving activities).

5.

a. Each organization, as appropriate, shall specify Section 6.5.5.1 action levels for determining the need for decontamination.

b. Each organization, as appropriate, shall establish Sections 6.6.2, 6.6.3, 6.6.4 the means for radiological decontamination of emergency personnel wounds, supplies, instruments and equipment, and for waste disposal.

6. Each licensee shall provide on-site contamination Sections 6.5.5.1, 6.5.5.2 control measures including:

a. Area access control;

b. Drinking water and food supplies;

c. Criteria for permitting return of areas and items to normal use (see Draft ANSI 13.12).

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7. Each licensee shall provide the capability for Sections 6.5.4, 6.6.2 decontaminating relocated on-site personnel, including Appendix E provisions for extra clothing and decontaminants suitable for the type of contamination expected, with particular attention given to radioiodine contamination of the skin.

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 )  L. MEDICAL AND PUBLIC HEALTH SUPPORT PLANNING STANDARD Arrangements are made for medical services for contaminated injured individuals. 1 1  The availability of an integrated Emergency Medical Services System and a public health emergency plan serving the area in which the facility is located and, as a minimum, equivalent to the Public Health Service Guide for Developing Health Disaster Plans, 1974, and to the requirements of an Emergency Medical Services System as outlined in the Emergency Medical Services System Act of 1973 (P.L. 93154 and amendments in 1979 P.L. 96142), should be a part of and consistent with overall State or local disaster control plans and should be compatible with the specific overall emergency response plan for the facility.

CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each organization shall arrange for local and backup Section 6.6.4 hospital and medical services having the capability for evaluation of radiation exposure and uptake, including assurance that persons providing these services are adequately prepared to handle contaminated individuals.

2. Each licensee shall provide for on-site first aid Section 6.6.2

~ capability.

3. Each State shall develop lists indicating the location Not Applicable to Licensee of public, private, and military hospitals, and other emergency medical services facilities within the State of contiguous States considered capable.of providing medical support for any contaminated injured individual. The listing shall include the name, location, type of facility, and capacity and any special radiological capabilities. These emergency medical services should be able to radiologically monitor contaminated personnel, and have facilities and trained personnel able to care for contaminated injured persons.

4. Each organization shall arrange for transporting Section 6.6.3 victims of radiological accidents to medical support facilities.

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)    M. RECOVERY AND RE-ENTRY PLANNING AND POST-ACCIDENT OPERATIONS PLANNING STANDARD General Plans for recovery and re-entry are developed.

CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each organization, as appropriate, shall develop Sections 9, 9.1, 9.2 general plans and procedures for re-entry and recovery, and describe the means by which decisions to relax protective measures (e.g., allow re-entry into an evacuated area) are reached. This process should consider both existing and potential conditions.

2. Each licensee plan shall contain the position/title, Sections 9, 9.1, 9.2 authority and responsibilities of individuals who will fill key positions in the facility recovery organization. This organization shall include technical personnel with responsibilities to develop, evaluate, and direct recovery and re-entry operations. The recovery organization recommended by the Atomic Industrial Forum's

'.:-) "Nuclear Power Plant Emergency Response Plan" dated October 11, 1979, is an acceptable framework.

3. Each licensee and State plan shall specify means for Section 9.2 informing members of the response organizations that a recovery operation is to be initiated, and of any changes in the organizational structure that may occur.

4. Each plan shall establish a method for periodically Section 9.3 estimating total population exposure.

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) N. EXERCISES AND DRILLS (*1)

PLANNING STANDARD Periodic exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. CROSS REFERENCE EVALUATION CRITERIA TO PLAN 1.

a. An exercise is an event that tests the integrated Sections 8.2, 8.2.1 capability and a major portion of the basic elements existing within emergency preparedness plans and organizations. The emergency preparedness exercise shall simulate an emergency that results in off-site radiological releases which would require response by off-site authorities. Exercises shall be conducted as set forth in NRG and FEMA rules.

b. An exercise shall include mobilization of State and Sections 8.2, 8.2.1 Local Personnel and Resources adequate to verify the capability to respond to an accident scenario requiring response. The organization shall provide for a critique of the annual exercise by Federal and State Observers/Evaluators. The scenario should be varied from year to year such that all major elements of the plans and preparedness organizations are tested within a five year period.

Each organization should make provisions to start an exercise between 6:00 p.m. and midnight, and another between midnight and 6:00 a.m. once every six years. Exercises should be conducted under various weather conditions. Some exercises should be unannounced. Notes:

             *1 - As supplemented by "Interim Staff Guidance - Emergency Planning for Nuclear Power Plants," NSIR/DPR-ISG-01, Rev. 0, November 2011.

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2. A drill is a supervised instruction period aimed at Sections 8.2, 8.2.3 testing, developing and maintaining skills in a particular operation. A drill is often a component of an exercise.

A drill shall be supervised and evaluated by a qualified drill instructor. Each organization shall conduct drills in addition to the annual exercise at the frequencies indicated below:

a. COMMUNICATION DRILLS Communications with State and Local Governments Section 8.2.3.5 within the plume exposure pathway Emergency Planning Zone shall be tested monthly.

Communications with Federal Emergency Response Organizations and States within the ingestion pathway shall be tested quarterly. Communications between the nuclear facility, State and Local Emergency Operations Centers, and Field Assessment Teams shall be tested annually. Communication drills shall also include the aspect of understanding the content of messages.

b. FIRE DRILLS Fire drills shall be conducted in accordance with the Section 8.2.3.1 plant (nuclear facility) technical specifications.

c. MEDICAL EMERGENCY DRILLS A medical emergency drill involving a simulated Section 8.2.3.2 contaminated individual which contains provisions for participation by the local support services agencies (i.e., ambulance and off-site medical treatment facility) shall be conducted annually. The off-site portions of the medical drill may be performed as part of the required annual exercise.

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d. RADIOLOGICAL MONITORING DRILLS Plant environs and radiological monitoring drills Section 8.2.3.3 (on-site and off-site) shall be conducted annually.

These drills shall include collection and analysis of all sample media (e.g., water, vegetation, soil, and air), and provisions for communications and record keeping. The State drills need not be at each site. Where appropriate, local organizations shall participate.

e. HEALTH PHYSICS DRILLS (1) Health Physics drills shall be conducted Section 8.2.3.4 semiannually which involve response to, and analysis of, simulated elevated airborne and liquid samples and direct radiation measurements in the environment. The State drills need not be at each site.

(2) Analysis of in-plant liquid samples with actual elevated radiation levels including use of the Post-Accident Sampling System shall be included in Health Physics drills by licensees annually.

3. Each organization shall describe how exercises and drills are to be carried out to allow free play for decision making and to meet the following objectives. Pending the development of exercise scenarios and exercise evaluation guidance by NRG and FEMA the scenarios for use in exercises and drills shall include, but not be limited to, the following:

a. The basic objective(s) of each drill and exercise and Sections 8.2.1, 8.2.3 appropriate evaluation criteria;

b. The date(s), time period, place(s) and participating Section 8.2.1 organizations;

c. The simulated events; Section 8.2.1 IEMERGENCY PLAN REVISION 76 PAGE 152 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

d. A time schedule of real and simulated initiating Section 8.2.1 events;

e. A narrative summary describing the conduct of the Section 8.2.1 exercises or drills to include such things as simulated casualties, off-site fire department assistance, rescue of personnel, use of protective clothing, deployment of Radiological Monitoring T earns, and public information activities; and

f. A description of the arrangements for and Section 8.2.1 advance materials to be provided to official observers.

4. Official observers from Federal, State or Local Section 8.2.1 Governments will observe, evaluate and critique the required exercises. A critique shall be scheduled at the conclusion of the exercise to evaluate the ability of organizations to respond as called for in the plan. The 0 critique shall be conducted as soon as practicable after the exercise, and a formal evaluation should result from the critique.

5. Each organization shall establish means for evaluating Section 8.2.1 observer and participant comments on areas needing improvement, including emergency plan procedural changes, and for assigning responsibility for implementing corrective actions. Each organization shall establish management control used to ensure that corrective actions are implemented.

I EMERGENCY PLAN REVISION 76 PAGE 153 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ,,___,_) 0. RADIOLOGICAL EMERGENCY RESPONSE TRAINING PLANNING STANDARD Radiological emergency response training is provided to those who may be called on to assist in an emergency. CROSS REFERENCE EVALUATION CRITERIA TO PLAN 1 . Each organization shall assure the training of Sections 8.1, 8.1.1 appropriate individuals.

a. Each facility to which the plant applies shall provide Section 8.1.3 site specific emergency response training for those off-site emergency organizations who may be called upon to provide assistance in the event of an emergency. 1 1

Training for hospital personnel, ambulance/rescue, police, and fire departments shall include the procedures for notification, basic radiation protection, and their expected roles. For those local services support organizations who will enter the site, training shall also include site access procedures and the identity (by position and title) of the individual in the 0 on-site emergency organization who will control the organizations' support activities. Off-site emergency response support personnel should be provided with appropriate identification cards where required.

b. Each off-site response organization shall participate Not Applicable to Licensee in and receive training. Where mutual aid agreements exist between local agencies such as fire, police, and ambulance/rescue, the training shall also be offered to the other departments who are members of the mutual aid district.

2. The training program for members of the on-site Sections 8.1.1, 8.1.2 emergency organization shall, besides classroom training, include practical drills in which each individual demonstrates ability to perform his assigned emergency function. During the practical drills, on-the-spot correction of erroneous performance shall be made and a demonstration of the proper performance offered by the instructor.

3. Training for individuals assigned to licensee First Aid Section 8.1.1 Teams shall include courses equivalent to Red Cross Multi-Media.

IEMERGENCY PLAN REVISION 76 PAGE 154 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE EVALUATION CRITERIA TO PLAN

4. Each organization shall establish a training program for instructing and qualifying personnel who will implement radiological emergency response plans. 2 The specialized initial training and periodic retraining programs (including the scope, nature and frequency) shall be provided in the following categories:

2 If State and Local Governments lack the capability and resources to accomplish this training, they may look to the licensee and the Federal government (FEMA) for assistance in this training.

    ""  NRG and FEMA encourage State and Local Governments which have these capabilities to continue to include them in their training programs.

a. Directors or coordinators of the response Section 8.1.1 organizations;

b. Personnel responsible for accident assessment; Section 8.1.1

c. Radiological Monitoring Teams and Radiological Section 8.1 .1 Analysis personnel;

d. Police, Security, and Fire-fighting personnel; Sections 8.1.1, 8.1.3

e. Repair and damage control/correctional action Section 8.1.1 teams (on-site);

f. First aid and rescue personnel; Section 8.1.1

g. Local support services personnel including Civil Section 8.1.3 Defense/Emergency Service personnel;

h. Medical support personnel; Sections 8.1.1, 8.1.3

i. Licensee's headquarters support personnel; Section 8.1.1

j. Personnel responsible for transmission of Section 8.1.1 emergency information and instructions.

5. Each organization shall provide for the initial and annual Section 8.1.1 retraining of personnel with emergency response responsibilities.

IEMERGENCY PLAN REVISION 76 PAGE 155 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) P. RESPONSIBILITY FOR THE PLANNING EFFORT: DEVELOPMENT, PERIODIC REVIEW AND DISTRIBUTION OF EMERGENCY PLANS PLANNING STANDARD Responsibilities for plan development and review and for distribution of emergency plans are established, and planners are properly trained. CROSS REFERENCE EVALUATION CRITERIA TO PLAN

1. Each organization shall provide for the training of Section 8.1.2 individuals responsible for the planning effort.

2. Each organization shall identify by title the individual Section Introduction, 8.3 with the overall authority and responsibility for radiological emergency response planning.

3. Each organization shall designate an Emergency Section 8.3 Planning Coordinator with responsibility for the development and updating of emergency plans and coordination of these plans with other response

,.)

\_

organizations.

4. Each organization shall update its plan and agreements, Section 8.5 as needed, review and certify it to be current on an annual basis. The update shall take into account changes identified by drills and exercises.

5. The emergency response plans and approved changes Section 8.5 to the plans shall be forwarded to all organizations and appropriate individuals with responsibility for implementation of the plans. Revised pages shall be dated and marked to show where changes have been made.

6. Each plan shall contain a detailed listing of supporting Section 2 plans and their source. Appendix F

7. Each plan shall contain as an appendix listing, by title, Appendix A procedures required to implement this plan. The listing shall include the section(s) of the plan to be implemented by each procedure.

\...J IEMERGENCY PLAN REVISION 76 PAGE 1 56 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1)

~                      EVALUATION CRITERIA CROSS REFERENCE TO PLAN

8. Each plan shall contain a specific table of contents. Table of Contents Plans submitted for review should be cross referenced Appendix B to these criteria.

9. Each licensee shall arrange for and conduct Section 8.5 independent reviews of the emergency preparedness program at least every 12 months. (An independent review is one conducted by any competent organization either internal or external to the licensee's organization, but who are not immediately responsible for the emergency preparedness program.) The review shall include the emergency plan, its implementing procedures and practices, training, readiness testing, equipment, and interfaces with State and Local Governments. Management controls shall be implemented for evaluation and correction of review findings. The results of the review, along with recommendations for improvements, shall be documented, reported to appropriate licensee corporate and plant management, and involved Federal, State,

~) and Local organizations, and retained for a period of five years.

10. Each organization shall provide for updating telephone Section 8.5 numbers in Emergency Procedures at least quarterly.

IEMERGENCY PLAN REVISION 76 PAGE 157 OF 192 j

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ~ CROSS-REFERENCE 10CFR50.47 NUREG-0654 1 Rev. 1 (FEMA REP 1) - Planning Standards b.1 A Assignment of Responsibility (Organization Control) b.2 B On-Site Emergency Organization b.3 C Emergency Response Support and Resources b.4 D Emergency Classification System b.5 E Notification Methods and Procedures b.6 F Emergency Communications b.7 G Public Education and Information b.8 H Emergency Facility and Equipment b.9 Accident Assessment b.10 J Protective Response b.11 K Radiological Exposure Control b.12 L Medical and Public Health Support b.13 M Recovery and Re-Entry Planning and Post-Accident Operations b.14 N Exercises and Drills 0 b.15 0 p Radiological Emergency Response Training Responsibility for the Planning Effort: b.16 Development, Periodic Review and Distribution of Emergency Plans 10CFR50.47.b contains sixteen standards to be met by a licensee's Emergency Plan. NUREG-0654/FEMA-REP-1 Rev. 1, contains sixteen planning standards which must be addressed in the licensee's Emergency Plan. The NUREG-0654, Rev. 1 (FEMA REP 1) Planning Standards are word for word duplications of the standards found in 10CFR50.47.b. The cross-reference between NUREG-0654, Rev. 1 (FEMA REP 1 ), and the NPPD Emergency Plan found in Appendix B of the NPPD Emergency Plan adequately provides a cross-reference to the standards in 10CFR50.47.b. IEMERGENCY PLAN REVISION 76 PAGE 158 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE 10CFR50 APPENDIX E TO CNS EMERGENCY PLAN NUREG-0654, Rev. 1 10CFR50, Appendix E.IV (FEMA REP 1) NPPD Emergency Plan Section A.1 A.1.b, c 5, 5.1 A.2a A.1.d, B.2, 3 5.1, 5.2 A.2b B.5 5.2 A.2c B.2, 3 5.2 A.3 8.7 5.3 8.5 5.2 A.4 I 4.2, 6.2.4, 6.2.5, 6.3, 6.5 A.5 B.8 5.3, Appendix D 5.4 A.6 B.9 6.5, 6.6, Appendices D, F 5.4.1, 5.4.2, 5.4.3, 5.4.4, 6.4 A.7 C 6.5, 6.6, Appendices D, F A.8 J.9 6.2.4, 6.5, 6.6, Appendix D () A.9 NIA 5.1, Figure 5.2-1 B.1 D, I, J 4, 6.3, 7.5, 8.1.3 C.1 D 4, 6. 7.4 C.2 D 4 5.4, 6, 6.2.4, 6.2.5, 6.5 D.1 E Appendix D D.2 G.1, 2 8.1.4 6.2.4, 6.5, Table 4.1-5, Table 4.1-6, D.3 E Table 4.1-7, Table 4.1-8 E.1 K.3a, b 6.5.5.1, 6.6.1, 7.2, Appendix E E.2 H.6, 7, 8, I 4, 7.5, 7.7 E.3 K.5, 6, 7 6.5, 6.6 E.4 L 6.6 E.5 L 6.6, Appendix D E.6 L 6.6.3, 6.6.4, Appendix D E.7 L 6.6.3, 6.6.4, Appendix D IEMERGENCY PLAN REVISION 76 PAGE 159 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) ~ CROSS REFERENCE 10CFR50 APPENDIX E TO CNS EMERGENCY PLAN NUREG-0654, Rev. 1 10CFR50, Appendix E.IV (FEMA REP 1) NPPD Emergency Plan Section E.8 H.1, 2 6.5, 7.2.1, 7.2.2, 7.2.3 E.9 F 7, 7.3, 8.2 F N, 0 8.1, 8.2, 8.4 G P.2, 3 8.3, 8.4, 8.5, 8.6 H M 9 NIA 6.5 IEMERGENCY PLAN REVISION 76 PAGE 160 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) } CROSS REFERENCE NEI 99-01, Revision 5 TO EAL NUMBER FOR NOTIFICATION OF UNUSUAL EVENT CONDITIONS COOPER NUCLEAR STATION NEI 99-01, Rev 5 EMERGENCY ACTION LEVEL INITIATING CONDITION NUMBER AU1, Example 1 AU1.1 AU1, Example 2 AU1.2 AU 1, Example 3 AU1.3 AU1, Example 4 Not Applicable AU1, Example 5 Not Applicable AU2, Example 1 AU2.1 AU2, Example 2 AU2.2 CU1, Example 1 CU2.1 CU2, Example 1 CU2.2 CU2, Example 2 CU2.3 CU3, Example 1 CU1.1 CU4, Example 1 CU3.1 CU4, Example 2 CU3.2 CU6, Examples 1, 2 CU4.1 CU7, Example 1 CU6.1 GUS, Example 1 CU5.1 GUS, Example 2 Not Applicable FU1, Example 1 FU1.1 HU1, Example 1 HU1.1 HU1, Example 2 HU1.2 HU1, Example 3 HU1.4 HU1, Example 4 HU1.3 HU1, Example 5 HU1.5 HU2, Example 1 HU2.1 HU2, Example 2 HU2.2 HU3, Example 1 HU3.1 HU3, Example 2 HU3.2 HU4, Examples 1, 2, 3 HU4.1 IEMERGENCY PLAN REVISION 76 PAGE 161 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1)

/,..)                                 CROSS REFERENCE NEI 99-01, Revision 5 TO EAL NUMBER FOR NOTIFICATION OF UNUSUAL EVENT CONDITIONS COOPER NUCLEAR STATION NEI 99-01, Rev 5 EMERGENCY ACTION LEVEL INITIATING CONDITION NUMBER HU5, Example 1                                          HU6.1 SU1, Example 1                                          SU1.1 SU2, Example 1                                          SU3.1 SU3, Example 1                                          SU4.1 SU4, Example 1                                          SU5.1 SU4, Example 2                                          SU5.2 SU5, Examples 1, 2                                        SU6.1 SU6, Examples 1, 2                                        SUB.1 SUB, Example 1                                          SU2.1 SUB, Example 2                                      Not Applicable E-HU1, Example 1                                         EU1.1 (J

IEMERGENCY PLAN REVISION 76 PAGE 162 OF 192 j

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE NEI 99-01, Revision 5 TO EAL NUMBER FOR ALERT CONDITIONS / COOPER NUCLEAR STATION NEI 99-01, Rev 5 EMERGENCY ACTION LEVEL INITIATING CONDITION NUMBER AA 1, Example 1 AA1.1 AA 1, Example 2 AA1.2 AA 1, Example 3 AA1.3 AA 1, Example 4 Not Applicable AA 1, Example 5 Not Applicable AA2, Example 2 AA2.1 AA2, Example 1 AA2.2 AA3, Example 1 AA3.1 CA3, Example 1 CA1.1 CA 1, Examples 1, 2 CA2.1 CA4, Example 1, 2 CA3.1

J FA1, Example 1 HA 1, Example 1 FA1.1 HA1.1 HA 1 , Example 2 HA1.2 HA 1 , Example 4 HA1.3 A 1, Example 3 HA1.4 HA 1 , Example 6 HA1.5 HA 1, Example 5 HA1.6 HA2, Example 1 HA2.1 HA3, Example 1 HA3.1 HA4, Examples 1, 2 HA4.1 HA5, Example 1 HA5.1 HA6, Example 1 HA6.1 SA5, Example 1 SA1.1 SA2, Example 1 SA2.1 SA4, Example 1 SA4.1 IEMERGENCY PLAN REVISION 76 PAGE 163 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE NEI 99-01, Revision 5 TO EAL NUMBER FOR SITE AREA EMERGENCY CONDITIONS COOPER NUCLEAR STATION NEI 99-01, Rev 5 EMERGENCY ACTION LEVEL INITIATING CONDITION NUMBER AS 1, Example 1 AS1.1 AS1, Example 2 AS1.2 AS1, Example 3 Not Applicable AS1, Example 4 AS1.3 CS1, Example 1 CS2.1 CS1, Example 2 CS2.2 CS1, Example 3 CS2.3 FS1, Example 1 FS1.1 HS4, Example 1 HS4.1 HS2, Example 1 HS5.1 HS3, Example'1 HS6.1

0

)

 -~        SS1, Example 1                                            S81.1 SS2, Example 1                                            SS2.1 SS6, Example 1                                            SS4.1 SS3, Example 1                                            SS?.1 IEMERGENCY PLAN                                  REVISION 76            PAGE 164 OF 192 I

APPENDIX B CROSS INDEX TO NUREG-0654, REV. 1 (FEMA REP 1) CROSS REFERENCE NEI 99-01, Revision 5 TO EAL NUMBER FOR GENERAL EMERGENCY CONDITIONS COOPER NUCLEAR STATION NEI 99-01, Rev 5 INITIATING CONDITION EMERGENCY ACTION LEVEL NUMBER AG1, Example 1 AG1.1 AG1, Example 2 AG1.2 AG1, Example 3 Not Applicable AG1, Example 4 AG1.3 CG1, Example 1 CG2.1 CG1, Example 2 CG2.2 FG1, Example 1 FG1.1 HG1, Examples 1, 2 HG4.1 HG2, Example 1 HG6.1 SG1, Example 1 SG1.1 SG2, Example 1 SG2.1

O IEMERGENCY PLAN REVISION 76 PAGE 165 OF 192 I

(_ J APPENDIX C EVACUATION ROUTES/MAPS APPENDIX C EVACUATION ROUTES/MAPS TO lrdmy 9 E MERGE NCY PLAN REVISION 76 PAGE 166 OF 192

l '-._/ ) APPENDIX C EVACUATION ROUTES/MAPS N E B R A S K A 10 E\MCUA11Clll ACUTES --.ceneia To Falls~. FALiLSCITY: us:~1o ... -..... ll'alloOii,llfddloSdoool

                                                               ~61....,bll.S.~                         1411, and Murton 71"-,lhioHlkCilJ.                       9,,,e6o.

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Ri,o,-QI-- EMERGENCY PLAN REVISION 76 PAGE 167 OF 192

APPENDIX C EVACUATION ROUTES/MAPS 10 Mile EPZ IE MERGENCY PLAN REVISION 76 PAGE 168 OF 192 j

l J APPENDIX C EVACUATION ROUTES/MAPS

f I 14

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            .,_--:, 2, s_,,10 Mfle Rln.11,s     o.t.:W,roili c¥ ~,1!1'r. lSI\IUJt 'tWi41 C*!io f(trt..,N.trin&.!"'P'O ..

EMERGENCY PLAN REVISION 76 PAGE 169 OF 192

APPENDIX C EVACUATION ROUTES/MAPS COOPER NUCEAR STATION-POPULATION ESTIMATES NW NE fi!*l [IE] WNW SfE 1ud ctw'.l: ESE (sa -1, SE [I[]

                                                                                                    *~-_; 10 Mila to [,Z Bovnclafy N

2015 Extrapolated Resident Population llllot ~bw ...... O*l 0

                                  ,0 .

1-l 1-l i Jei1 - ,,, JD 3- 4 m 4Sl w

                                              ~       -

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J.44 U2' Flcurct 4. ~rmanctnt .EPZ Resldctnt Population by .Sector Cooper Nuclec1r Statlon l3 KLO Engineering, J>.C.

Populatlon Update Analysis - 2015 Rev. O IEMERGENCY PLAN R EVISION 76 PAGE 170 OF 192 I

APPENDIX C EVACUATION ROUTES/MAPS PRE-SELECTED SAMPLING POINTS ) EMERGENCY PLAN REVISION 76 PAGE 171 OF 192

APPENDIX C EVACUATION ROUTES/MAPS 2C10 Z0 15 C,tr dp0Lit<'d Arc,, P0:Juht1o n Po:1-..: 1tio n

                                                                                                                 . 1 282 1=Mlle Reglon .'i                                                     
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                                        ~           .Total:       J.

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w 3,668 ..3 616 *-~ - ... 201 . -- 201 ... W~)N NW 63.... 63.

                                    *-NNW .                             18 Cooper Nudear Station                                                12                                  KLD Engineering, P.C.

Population Update Analysls - 2015 Rev.O I EMERGENCY PLAN REVISION 76 PAGE 172 OF 192 I

l ) APPENDIX C EVACUATION ROUTES/MAPS Table 7-1. Time to Clear the Indicated Area of $1. Percent.l l~the Affected Population Summer 5..ummer Summer ; E MERGE NCY PLAN REVI SION 76 PAGE 173 OF 192

APPENDIX C EVACUATION ROUTES/MAPS Area(s) Shelter__-in-Place IEMERGENCY PLAN REVISION 76 PAGE 17 4 OF 192 I

APPENDIX D LETTERS OF AGREEMENT APPENDIX D LETTERS OF AGREEMENT Letters of Agreement (LOAs) supporting the CNS Emergency Plan are listed on the following page, and are incorporated in the Emergency Plan by reference. The Emergency Plan signature page verifies a signed copy of a current Letter of Agreement is on file per this Appendix. Copies of the current Letters of Agreement and a listing of their effective dates are maintained in the Emergency Preparedness office. Letters of Agreement are reviewed annually by the CNS Emergency Preparedness Department. Each organization is then contacted. The type of support defined in the letters is discussed to determine if any significant changes have occurred. If significant changes have occurred, a Letter of Agreement is requested from the agency. If there are no significant changes, the Letters of Agreement are certified current by the CNS Emergency Preparedness Department and documented ,With a record of telephone conversation or other appropriate documentation. A change in original signatory(ies) to a given Letter of Agreement does not in itself require revision of that Letter. The documentation associated with this review process is maintained by the CNS Emergency Preparedness Department. 0 IEMERGENCY PLAN REVISION 76 - PAGE 175 OF 192 I

APPENDIX D LETTERS OF AGREEMENT LETTERS OF AGREEMENT Letters of Agreement supporting the CNS Emergency Plan are certified annually. Copies of the current Letters of Agreement and listing of their effective dates are maintained in the Emergency Preparedness office. Agreement 1 . Nemaha County Hospital

2. Auburn Rescue Squad

3. Nebraska State Patrol

4. Nebraska State Patrol/Nebraska State Emergency Management Agency

5. Nebraska State Emergency Management Agency/NPPD 6.

6. Missouri State Emergency Management Agency

7. Kansas Division of Emergency Management 0 8. Iowa Emergency Management Division

9. Atchison County Commission

10. Nemaha County Commission

11. Richardson County Commission

12. Institute of Nuclear Power Operations

13. General Electric - Hitachi (SIL 324, Rev 7)

14. Omaha Public Power District/Fort Calhoun Station

15. Nebraska State Emergency Management Agency/

Nebraska Game and Parks Commission/NPPD

16. Peru State College

17. Brownville Fire Department

18. Auburn Fire Department

19. Nemaha Fire Department

20. Peru Fire Department IEMERGENCY PLAN REVISION 76 PAGE 176 OF 192 I

APPENDIX D LETTERS OF AGREEMENT r) 21. Nebraska City Volunteer Fire Department (NRC Commitment NLS2005104-04)

22. University of Nebraska Medical Center

23. Mirian

24. Midwest Medical Transport

25. Nemaha County Hospital (Use of helicopter pad)

(NRC Commitment NLS2012048-03)

26. Nebraska Emergency Management Agency (NEMA), Nebraska Department of Health and Human Services, Division of Public Health, and NPPD (NRC Commitment NLS2012048-03)

27. Nemaha Rescue Squad

28. Otoe County Emergency Management Agency IEMERGENCY PLAN REVISION 76 PAGE 177 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES This appendix contains a list of supplies that is typical of the inventory kept in the Emergency Response Lockers. For a precise inventory of the equipment and supplies, and who is responsible for it, refer to the most current revision of EPIP 5.7.21, Maintaining Emergency Preparedness-Emergency Exercises, Drills, Tests, and Evaluations. IEMERGENCY PLAN REVISION 76 PAGE 178 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - CONTROL ROOM A. GENERAL SUPPLIES

1. Coveralls

2. Shoe Covers

3. Gloves

4. Geiger-Mueller Survey Meter (Range 0-50 mR/Hr)

5. Ion-Chamber Survey Instrument (Range 0-50 R/Hr)

6. Dosimeter, Direct Reading, Electronic

7. Radiation Monitor (Frisker)

8. Step-Off Pad

9. Thyroid Blocking Tablets (Kl)

B. RESPIRATORY PROTECTION EQUIPMENT

~) NOTE - The air breathing equipment is not within the Emergency Locker. The cases are

,   located near the Emergency Locker for convenience, inspection, and maintenance.

1. Air Breathing Masks (Self-Contained with Voice Communicators)

2. Full-Face Filter Masks with Filters

3. Full-Face Filter Masks with Filters and Voice Communicators

4. Spare Air Cylinders C. MISCELLANEOUS (Supplies)

1. Plastic Bag, Large

2. Radiation Warning Signs

3. Radiation Barrier Rope

4. Radiation Warning Tape

5. Hand Lantern, with 6-Volt Battery

6. Flashlight, with Two "D" Cell Batteries

7. Batteries for Hand Lantern (6 Volt)

8. Batteries for Flashlights ("D" Cell)

9. Batteries for Mask Voice Communicators (9 Volt)

10. First Aid Kit IEMERGENCY PLAN REVISION 76 PAGE 1 79 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED -TSC/OSC A. GENERAL SUPPLIES AND PROTECTION EQUIPMENT

1. Flashlight, with Two "D" Cell Batteries

2. Masking Tape

3. Particulate Filter, 2"

4. Charcoal and Silver Zeolite Cartridges

5. Air Sample Plastic Bags and Labels

6. Smear Books

7. Spare Batteries

8. Personnel Radiation Monitor

9. Step-Off Pads

10. Protective Clothing (Full Sets)

11. Self-Contained Breathing Apparatus

12. Spare Bottle for SCBA

J 13. Thyroid Blocking Tablets (Kl)

14. Survey Instrument Ion Chamber (Range O to 50 R/hr)

15. !AC/Electrical Tool Kits

16. Volt Ohmmeter

17. Radiological Posting Supplies

18. Mechanical Maintenance Tool Kit

19. Coveralls, Paper

20. Shoe Covers, Disposable 14"

21. Gloves, Disposable

23. Continuous Air Monitor B. EMERGENCY RESCUE LOCKER EQUIPMENT

1. Wrecking Bar

2. Bolt Cutters

3. Hacksaw and Blades

4. Come-Along

-.__,) 5. Cable Slings I EMERGENCY PLAN REVISION 76 PAGE 180 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED -TSC/OSC

6. Hydraulic Jacks

7. Sledge Hammers

8. Porta Power

9. Web Slings

11. Safety Belt and Line

12. Fire Axe

13. Crow Bar

14. 200'-3-Part Block and Tackle

15. Battery Lanterns IEMERGENCY PLAN REVISION 76 PAGE 181 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - ALTERNATE OSC (AOSC) A. GENERAL SUPPLIES AND PROTECTION EQUIPMENT

1. Coveralls, Paper

2. Shoe Covers

3. Gloves, Disposable

4. Step-Off Pad

5. Area Radiation Monitors

6. Continuous Air Monitor

7. Radiation Monitor (Frisker)

8. Flashlight, with 2 "D" Cell Batteries

9. Spare Batteries ("D" Cell)

10. Thyroid Blocking Tablets (Kl)

11. Team Dispatch Forms

O IEMERGENCY PLAN REVISION 76 PAGE 182 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - EOF A. GENERAL SUPPLIES

1. Coveralls

2. Shoe Covers

3. Gloves

4. Extendable Probe Survey Instrument (Range 0-1,000 R/Hr)

5. Ion-Chamber Survey Meter (Range 0-50 R/Hr)

6. Geiger-Mueller Survey Meter (Range 0-50 mR/Hr)

7. Sample Holder with Pancake-Type Detector

8. Scaler Electronic Package

9. Dosimeter, Direct Reading Electronic

10. Thyroid Blocking Tablets

11. Spare Batteries ("AA" Cell)

12. Charcoal Filter for Air Samplers

)

,, 13. Silver Zeolite Cartridges for Air Samplers

14. Extension Cord, Electric (50')

15. Radiation Monitor (Frisker)

16. Area Radiation Monitor

17. Continuous Air Monitor

18. Step-Off Pads B. MISCELLANEOUS (Supplies)

1. Plastic Sheeting

2. Plastic Bag, Small

3. Plastic Bag, Large

4. Radiation Warning Signs

5. Radiation Barrier Rope

6. Smear Books

7. Radiation Warning Tape

8. Hand Lantern with 6-Volt Battery 0 9. Flashlight, with Two "D" Cell Batteries IEMERGENCY PLAN REVISION 76 PAGE 183 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES

10. Batteries for Hand Lantern (6 Volt)

11. Batteries for Flashlights ("D" Cell)

12. Small Hand Tool Kit with Straight Slot Screwdriver, Phillips Screwdriver, Small Pliers; and Small Vise Grip C. FIRST AID AND RESCUE EQUIPMENT NOTE - Stretcher stored near Emergency Locker.

1. First Aid Kit

2. Stretcher 0

IEMERGENCY PLAN REVISION 76 PAGE 184 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - WEST WAREHOUSE A. EMERGENCY FIELD MONITORING KIT SUPPLIES

1. Full-Face Respirator

2. Hand Lantern

3. Spare Batteries

4. Dosimeter, Direct Reading Electronic

5. Health Physics Procedure 9.EPIN.1, Emergency Air Samplers

6. Calculator

7. Portable Radios

8. Geiger Mueller Survey Instrument

9. Ion Chamber Survey Instrument

10. Paper Coveralls

11. Rubber Shoe Covers 12 . Sample Bottles

.() 13. Masslin Cloths '-~

14. One-Piece Plastic Coveralls

15. Complete Set of EPIPs

16. 2" Air Sample Filters

17. Silver Zeolite Cartridges

18. Charcoal Cartridges

19. 2" Millipore Air Sample Filters

20. Smear Books

21. Air Sampler with Head

22. Radioactive Material Stickers

23. 10-Mile Radius Map

24. Site Map

25. Plastic Bags

26. Disposable Gloves

27. Thyroid Blocking Tablets (Kl)

28. Combination Cartridge for Respirator

29. Shovel IEMERGENCY PLAN REVISION 76 PAGE 185 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - WEST WAREHOUSE

30. Masking Tape

31. Plastic Sheeting

32. Plastic Pipet

33. 2 cc Vial

34. Sample Labels

35. Grass Shears

36. Bolt cutters IEMERGENCY PLAN REVISION 76 PAGE 186 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - COMMUNICATIONS BUILDING A. PERSONNEL DECONTAMINATION SUPPLIES

1. Soap

2. Septisol (Germicide)

3. Lanolin

4. Swabs, Cotton Tipped, 100s

5. Compresses, Gauze, 3" x 3", 100s

6. Towels, Paper

7. Beaker, Plastic, 100 ml

8. Hand Brush

9. Towels B. FIRST AID AND RESCUE EQUIPMENT NOTE - Stretcher stored near Emergency Locker.

1. First Aid Kit

2. Stretcher

3. Rope, 1/2"-50' IEMERGENCY PLAN REVISION 76 PAGE 187 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED-,AMBULANCE A. EMERGENCY EQUIPMENT MAINTAINED FOR AMBULANCE

1. Dosimeter, Direct Reading Electronic

2. Spare Batteries

3. DLR Badge

4. Geiger-Mueller Survey Meter

5. Ion-Chamber Survey Instrument

6. Radiation Tags

7. Smear Books

8. DLR Badging Record C)

IEMERGENCY PLAN REVISION 76 PAGE 1 88 OF 192 j

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - HOSPITAL A. EMERGENCY EQUIPMENT MAINTAINED AT HOSPITAL

1. Radiation Barrier Rope

2. Masking Tape

3. Absorbent Paper

4. Plastic Sheeting

5. Applicable Radiation Warning Signs

6. Shoe Covers

7. Bags, Plastic (Large)

8. Bags, Plastic (Small)

9. Radiation Marking Tape

10. Coveralls

11. Gloves, Rubber Disposable

12. Cardboard Boxes, 2' x 3'

() 13. Masolin Cloths

14. Step-Off Pad u

I EMERGENCY PLAN REVISION 76 PAGE 189 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY EQUIPMENT MAINTAINED - OFF-SITE ASSEMBLY AREA A. EMERGENCY EQUIPMENT MAINTAINED AT OFF-SITE ASSEMBLY AREA

1. Geiger-Mueller Survey Instrument

2. Disposable Coveralls

3. Disposable Gloves

4. 2" Masking Tape

5. Large Poly Bags

6. Small Poly Bags

7. Bar Soap

8. Bath Towel

9. Procedure 9.EN-RP-104, Personnel Contamination

10. Pumice Soap

11. Lanolin

12. Swabs, Cotton-Tipped

() 13. Paper Towels

14. Hand Brush, Soft Bristle

15. High-Top Bootie

16. Shoe Cover(s) Various Sizes

17. Rad Rope

18. Radiological Posting Signs With Inserts IEMERGENCY PLAN REVISION 76 PAGE 190 OF 192 I

APPENDIX E LISTING OF EMERGENCY KITS AND GENERAL CATEGORIES OF PROTECTIVE EQUIPMENT AND SUPPLIES FOR EMERGENCY PURPOSES EMERGENCY VEHICLES MAINTAINED - CNS A. EMERGENCY VEHICLES MAINTAINED AT CNS

1. All Wheel/Four Wheel Drive Vehicle with High and Low Band Radio for Emergency Preparedness/Security Use Only

2. All Wheel/Four Wheel Drive Vehicle with High and Low Band Radio for Emergency Preparedness/Security Use Only

3. Ambulance (2WD), Chassis with Medical Configuration and Two-Way Radio for Medical Use Only.

IEMERGENCY PL.AN REVISION 76 PAGE 191 OF 192 j

APPENDIX F INTERFACING EMERGENC Y PLANS APPEt-0IX F INTERFACING EMERGENCY PLANS NEBRASKA

1. State of Nebraska Radiological Emergency Response Plan for Nuclear Power Plant incidents-Nebraska Emergency Management Agency.

2. Radiological Emergency Response Plan for Nuclear Power Plant incidents for Richardson County-Nebraska Emergency Management Agency, Richardson County Emergency Management Agency.

3. Radiological Emergency Response Plan for Nuclear Power Plant incidents for Nemaha County-Nebraska Emergency Management Agency, Nemaha County Emergency Managemen t Agency.

4. Radiological Emergency Reception Plan for Nuclear Power Plant incidents for Otoe County-Nebraska Emergency Management Agency, Otoe County Emergency Managemen t Agency.

MISSOURI

5. State of Missouri State Emergency Management Agency Nuclear Accident Plan-Missouri State Emergency Managemen t Agency.

C) 6. Atchison County Radiological Emergency Response Plan-Atchison County, Missouri. KANSAS

7. The State of Kansas, Radiological Emergency Response Plan for Nuclear Facilities -

Kansas Division of Emergency Management. IOWA

8. State of Iowa Radiological Emergency Response Plan Iowa Department of Homeland Security, Emergency Managemen t Division.

FEDERAL

9. National Response Framework-United States Nuclear Regulatory Commission; Department of Homeland Security, Federal Emergency Management Agency.

I IEMERGENCY PLAN REVISION 76 PAGE 192 OF 192 I

NLS2021001 Page 1 of343 ENCLOSURE2 Cooper Nuclear Station Emergency Plan Implementing Procedure 5.7.1, Emergency Classification, Revision 67

COOPER NUCLEAR STATION Operations Manual Emergency Preparedness EMERGENCY PLAN IMPLEMENTING PROCEDURE 5.7.1 EMERGENCY CLASSIFICATION Level of Use: MULTIPLE Quality: QAPD RELATED Effective Date: 12/16/20 Approval Authority: ITR-RDM Procedure Owner: EMERG PREP DRILL SCENARIO COORD PROCEDURE 5.7.1 REVISION 67 PAGE 1 OF 342

TABLE OF CONTENTS

1. ENTRY CONDffiONS [REFERENCE USE] .................................................... 3

2. INSTRUCTIONS [REFERENCE USE] ........................................................... 3 ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE

[INFORMATION USE] .................................................... 4 ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] .................................................. 16 ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] ................................................ 258 ATTACHMENT 4 EAL CLASSIFICATION MATRIX [INFORMATION USE] ..... 312 ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] ................................................ 314 ATTACHMENT 6 EAL DEFINITIONS AND ACRONYMS [INFORMATION USE] ....................................................................... 325 ATTACHMENT 7 INFORMATION SHEET [INFORMATION USE] ................. 337 ATTACHMENT 8 MATRIX BASIS CROSS-REFERENCE [INFORMATION USE] ....................................................................... 340 PROCEDURE 5.7.1 REVISION 67 PAGE 2 OF 342

1. ENTRY CONDITIONS [REFERENCE USE]

1.1 An Emergency Operation Procedure has been initiated; or 1.2 An unusual occurrence has taken place at or near site.

2. INSTRUCTIONS [REFERENCE USE]

2.1 PERFORM classification and declaration activities per EPIP 5.7.1.1 as supported by attachments in this procedure. PROCEDURE 5.7.1 REVISION 67 PAGE 3 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] ATTACHMENT 1 EAL SQ-IEME EXPLANATION AND RATIONALE [INFORMATION USE]

1. PURPOSE 1.1 This attachment along with Attachments 2, 3, 5, 6, and 7 provide an explanation and rationale for each Emergency Action Level (EAL)

Included In the EAL Scheme for Cooper Nuclear Station (CNS). It should be used in conjunction with EPIP 5. 7.1.1 to facilitate review of the CNS EALs, provide historical documentation for future reference, and serve as a resource for training. Decision-makers responsible for Implementation of the EAL Scheme may use these attachments as a technical reference in support of EAL interpretation. 1.2 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. 1.3 This procedure, its attachments, and the EAL Classification Matrix are controlled pursuant to 10CFR50.54(q).

2. DISCUSSION

2.1 BACKGROUND

2.1.1 An EAL is a pre-determined, site specific, observable threshold for a plant Initiating Condition (IC) that places the plant in a given Emergency Classification Level (ECL). An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (on-site or off-site); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency classification level. EALs are utilized to classify emergency conditions defined in the CNS Emergency Plan. 2.1.2 In 1992, the NRC endorsed NUMARC/NESP-007, Methodology for pevelopment of Emergency Action Levels, as an alternative to NUREG-0654 EAL guidance. PROCEDURE 5. 7 .1 REVISION 67 PAGE 4 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.1.3 NEI 99-01 (NUMARC/NESP-007), Revision 5, was the most recently accepted methodology for development of Emergency Action Levels approved by the NRC when CNS upgraded this classification scheme. 2.1.4 Using NEI 99-01, Revision 5, CNS conducted an EAL implementation upgrade project that produced the EALs discussed in this procedure. CNS has matched plant features and safety system designs that are unique to CNS to the generic NEI 99-01, Revision 5, guidance. Should CNS implement any design differences not addressed in NEI 99-01, Revision 5, CNS will have to consider the applicable failure mechanisms involved when assessing the impact on the site specific EALs. 2.2 FISSION PRODUCT BARRIERS 2.2.1 Many of the EALs derived from the NEI methodology are fission product barrier based. That is, the conditions that define the EALs are based upon 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. "Loss" means the barrier no longer assures containment of radioactive materials; "potential loss" infers an increased probability of barrier loss and decreased certainty of maintaining the barrier. 2.2.2 The primary fission product barriers are: 2.2.2.1 Fuel Clad (FC): The Fuel Clad barrier consists of the zircaloy fuel bundle tubes that contain the fuel pellets. 2.2.2.2 Reactor Coolant System (RCS): The RCS barrier is the Reactor Coolant System pressure boundary and includes the reactor vessel and all Reactor Coolant System piping up to the isolation valves. 2.2.2.3 Containment (PC): The Primary Containment barrier includes the drywell, the wetwell (torus), their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. PROCEDURE 5. 7 .1 REVISION 67 PAGE 5 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.3 EMERGENCY CLASSIFICATION BASED ON FISSION PRODUCT BARRIER DEGRADATION 2.3.1 The following criteria are the bases for event classification related to fission product barrier loss or potential loss: 2.3.1.1 Unusual Event: Any loss or any potential loss of Primary Containment. 2.3.1.2 Alert: Any loss or any potential loss of either Fuel Clad or RCS. 2.3.1.3 Site Area Emergency: Loss or potential loss of any two barriers. 2.3.1.4 General Emergency: Loss of any two barriers and loss or potential loss of third barrier. 2.4 EAL RELATIONSHIP TO EOPS 2.4.1 Where possible, the EALs have been made consistent with and utilize the conditions defined In the CNS Emergency Operating Procedures (EOPs). While the symptoms that drive Operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification, they define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. When these symptoms are clearly representative of one of the NE! Initiating Conditions, they have been utilized as an EAL. This permits rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs. PROCEDURE 5. 7 .1 REVISION 67 PAGE 6 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.5 SYMPTOM-BASED VS. EVENT-BASED APPROACH 2.5.1 To the extent possible, the EALs are symptom-based. That is, the action level threshold is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. However, a purely symptom-based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no pre-determined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized. 2.6 EAL ORGANIZATION 2.6.1 The CNS EAL scheme includes the following features: 2.6.1.1 Division of the EAL set into three broad groups.

a. GROUPS

1. EALs applicable under all plant operating modes - this group would be reviewed by the EAL-user any time emergency classification is considered.

2. EALs applicable only under MODE 1, 2, or 3 - this group would only be reviewed by the EAL-user when the plant is rn Hot Shutdown, Startup, or Power Operation mode.

3. EALs applicable only under MODE 4, 5, or Defueled - this group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling, or Defueled Mode.

PROCEDURE 5. 7 .1 REVISION 67 PAGE 7 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE]

b. The purpose of the groups is to avoid review of EALs that cannot be applicable in the current operating mode of the plant. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden, and thereby, speeds identification of the EAL that applies to the emergency.

2.6.1.2 Within each of the above three groups, assignment of EALs to categories/subcategories.

a. Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. Subcategories are used, as necessary, to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The CNS EAL categories/subcategories are contained in Attachment 5.

2.6.1.3 The primary tool for determining the emergency classification level is the EAL classification matrix. The user of the EAL classification matrix may (but is not required to) consult the EAL Technical Bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Sections 2.7 and 2.7.1.11 of this attachment, and Attachments 2 and 3 for such information.

2. 7 TECHNICAL BASES INFORMATION

2. 7.1 EAL Technical Bases are provided in Attachment 2 for each EAL according to EAL group, EAL category (A, C, H, S, E, and F), 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:

2.7.1.1 CATEGORY LETTER AND TITLE. 2.7.1.2 SUBCATEGORY NUMBER AND TITLE. 2.7.1.3 INITIATING CONDITION (IC). PROCEDURE 5. 7 .1 REVISION 67 PAGE 8 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE]

2. 7 .1.4 SITE-SPECIFIC DESCRIPTION OF THE GENERIC IC GIVEN IN NEI 99-01.

2.7.1.5 EAL IDENTIFIER (ENCLOSED IN RECTANGLE)

a. Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to on-site and off-site personnel. Four characters define each EAL identifier:

1. First character (letter): Corresponds to the EAL category as described in Attachment 5 (A, C, H, S, E, or F).

2. Second character (letter): The emergency classification (G, S, A, or U).

3. Third character (number): Subcategory number within the given category.

a) 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.

a) If subcategory has only one EAL, it is given the number one (1).

2. 7.1.6 CLASSIFICATION (ENCLOSED IN RECTANGLE)

a. Unusual Event (U), Alert (A), Site Area Emergency (S), or General Emergency (G).

2. 7 .1. 7 EAL (ENCLOSED IN RECTANGLE)

a. Exact wording of the EAL as it appears in the EAL classification matrix.

PROCEDURE 5.7.1 REVISION 67 PAGE 9 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.7.1.8 MODE APPLICABILITY

a. One or more of the following plant operating conditions comprise the MODE to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled (DEF), All or N/A - Not Applicable (see Section 2.7.1.11 for operating Mode definitions.)

2.7.1.9 CNS BASIS

a. Provides CNS-relevant information concerning the EAL.

2.7.1.10 CNS BASIS REFERENCE(S)

a. Site-specific source documentation from which the EAL ls derived.

2. 7 .1.11 NEI 99-01 BASIS

a. Provides a description of the rationale for the EAL as provided in NEI 99-01.

2.8 OPERATING MODE APPLICABILITY 2.8.1 MODES 2.8.1.1 POWER OPERATION

a. Reactor mode switch is in RUN.

2.8.1.2 STARTUP

a. The mode switch is in either REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY.

2.8.1.3 HOT SHUTDOWN

a. The mode switch is in SHUTDOWN with all reactor vessel head closure bolts fully tensioned and reactor coolant temperature > 212°F.

PROCEDURE 5. 7 .1 REVISION 67 PAGE 10 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.8.1.4 COLD SHUTDOWN

a. The mode switch is in SHUTDOWN with all reactor vessel head closure bolts fully tensioned and reactor coolant temperature~ 212°F.

2.8.1.5 REFUELING

a. The mode switch is in either REFUEL or SHUTDOWN with one or more reactor vessel head closure bolts less than fully tensioned.

2.8.1.6 DEFUELED

a. All reactor fuel removed from reactor pressure vessel (full core off load during refueling or extended outage). This mode is designated as DEF in the EAL Classification Matrix.

2.8.2 The plant operating mode that exists at the time the event occurs (prior to any protective system or Operator action is 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. 2.8.3 For events that occur in Cold Shutdown or Refueling, escalation is per EALs that have Cold Shutdown or Refueling for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher. 2.9 VALIDATION OF INDICATIONS, REPORTS, AND CONDITIONS

2. 9.1 All classifications are to be based upon valid indications, reports, or conditions. Indications, reports, or conditions are considered valid when they are verified by (1) an instrument channel check, or (2) indications on related or redundant indications, or (3) by direct observation by plant personnel, such that doubt, related to the indication's operability, the condition's existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

PROCEDURE 5. 7.1 REVISION 67 PAGE 11 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.10 PLANNED VS. UNPLANNED EVENTS 2.10.1 Planned evolutions involve preplanning to address the limitations imposed by the condition, the performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition in accordance with the specific requirements of the CNS Technical Specifications. Activities which cause operation beyond that allowed by Technical Specifications, planned or unplanned, may result in an EAL threshold being met or exceeded. Planned evolutions to test, manipulate, repair, perform maintenance, or modifications to systems and equipment that result in an EAL value being met or exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the operating license. However, these conditions may be subject to the reporting requirements of 10CFRS0. 72. 2.11 CLASSIFYING TRANSIENT EVENTS 2.11.1 For some events, the condition may be corrected before a declaration has been made. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined; in other situations, further analyses (e.g., coolant radiochemistry sampling) may be necessary. Classify the event as indicated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met. 2.11.2 Existing guidance for classifying transient events addresses the period of time of event recognition and classification (15 minutes). However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response or

               *result from appropriate Operator actions.

2.11.3 There may be cases in which a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review) and the condition no longer exists. In these cases, an emergency should not be declared. PROCEDURE 5. 7 .1 REVISION 67 PAGE 12 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.11.4 Reporting requirements of 10CFR50. 72 are applicable and the guidance of NUREG-1022, Event Reporting Guidelines 10CFR50. 72 and 50.73, should be applied. 2.12 MULTIPLE SIMULTANEOUS EVENTS AND IMMINENT EAL THRESHOLDS 2.12.1 When multiple simultaneous events occur, the emergency classification level is based on the highest EAL reached. For example, two Alerts remain in the Alert category; or, an Alert and a Site Area Emergency is a Site Area Emergency. Further guidance is provided in RIS 2007-02, Clarification of NRC Guidance for Emergency Notifications during Quickly Changing Events. 2.12.2 Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classification levels (the early classification may permit more effective implementation of protective measures), it is nonetheless applicable to all emergency classification levels. 2.13 EMERGENCY CLASSIFICATION LEVEL DOWNGRADING 2.13.1 Another important aspect of usable EAL guidance is the consideration of what to do when the risk posed by an emergency is clearly decreasing. A combination approach involving recovery from General Emergencies and some Site Area Emergencies and termination from Unusual Events, Alerts, and certain Site Area Emergencies causing no long-term plant damage appears to be the best choice. Downgrading to lower emergency classification levels adds notifications but may have merit under certain circumstances. PROCEDURE 5.7.1 REVISION 67 PAGE 13 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] 2.14 CNS-TO-NEI 99-01 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a CNS EAL within the NEI 99-01 IC/EAL identification scheme. CNS N El 99-01, REV 5 CNS N El 99-01, REV 5 EXAMPLE EXAMPLE EAL IC EAL EAL IC EAL AUl.1 AUl 1 AUl.2 AUl 2 AUl.3 AUl 3 AU2.l AU2 1 AU2.2 AU2 2 AAl.1 AAl 1 AAl.2 AAl 2 AAl.3 AAl 3 AA2.1 AA2 2 AA2.2 AA2 1 AA3.1 AA3 1 ASl.1 ASl 1 ASl.2 ASl 2 ASl.3 ASl 4 AGl.1 AGl 1 AGl.2 AGl 2 AGl.3 AGl 4 CUl.1 CU3 1 CU2.1 CUl 1 CU2.2 CU2 1 CU2.3 CU2 2 CU3.1 CU4 1 CU3.2 CU4 2 CU4.1 CU6 1, 2 CUS.l CUB 1 CU6.l CU7 1 CAl.1 CA3 1 CA2.1 CAl 1, 2 CA3.l CA4 1, 2 CS2.l CSl 1 CS2.2 CSl 2 CS2.3 CSl 3 CG2.1 CGl 1 CG2.2 CGl 2 EUl.1 E-HUl 1 FUl.1 FUl 1 FAl.1 FAl 1 FSl.1 FSl 1 FGl.1 FGl 1 PROCEDURE 5.7.1 REVISION 67 PAGE 14 OF 342

ATTACHMENT 1 EAL SCHEME EXPLANATION AND RATIONALE [INFORMATION USE] CNS N EI 99-01, REV 5 CNS NEI 99-01, REV 5 EXAMPLE EXAMPLE EAL IC EAL EAL IC EAL HUl.1 HUl 1 HUl.2 HUl 2 HUl.3 HUl 4 HUl.4 HUl 3 HUl.S HUl s HU2.1 HU2 1 HU2.2 HU2 2 HU3.1 HU3 1 HU3.2 HU3 2 HU4.1 HU4 1, 2, 3 HU6.1 HUS 1 HAl.1 HAl HAl.1 HAl.2 HAl 2 HAl.3 HAl 4 HAl.4 HAl 3 HAl.S HAl 6 HAl.6 HAl s HA2.1 HA2 1 HA3.1 HA3 1 HA4.1 HA4 1, 2 HAS.1 HAS 1 HA6.1 HA6 1 HS4.1 HS4 1 HSS.l HS2 1 HS6.1 HS3 1 HG4.1 HGl 1, 2 HG6.1 HG2 1 SUl.1 SUl 1 SU2.1 SUB 1 SU3.1 SU2 1 SU4.1 SU3 1 SUS.1 SU4 1 SUS.2 SU4 2 SU6.l SUS 1, 2 SAl.1 SAS 1 SA2.1 SA2 1 SA4.l SA4 1 SSl.l SSl 1 SS2.1 SS2 1 SS4.l SS6 1 SS7.1 SS3 1 SU8.1 SU6 1, 2 SGl.1 SGl 1 SG2.1 SG2 1 PROCEDURE 5. 7.1 REVISION 67 PAGE 15 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] ATTACHMENT 2 EMERGENCY ACTION LEVEL TEOINICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than two times the ODAM limits for 60 minutes or longer EAL: AUl.1 Unusual Event Any valid gaseous monitor reading > Table A-1 column "UE" for~ 60 min. (NOTE 2) (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 16 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NOTE 2 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume the release duration has exceeded the applicable time_ if an_ on-going_ release _is detected_ and _the_ release start _time _is _unknown. _________ _ Table A-1 Effluent Monitor Classification Thresholds GE SAE ALERT UE Monitor for:?: 15 min. for~ 15 min. for~ 15 min. for~ 60 min. ERP 3.50E+08 µCl/sec 3.50E+07 µCl/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec Rx Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCl/sec 5.45E+05 µCi/sec 8.48E+04 µCi/sec (/'J 0 w Turb Bldg Vent 3.50E+07 µCl/sec 3.50E+06 µCl/sec 5.62E+05 µCi/sec 9.02E+04 µCi/sec (/'J c3 RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec C 200 x calculated 5 Rad Waste Effluent - - alarm values* 2 x calculated 0

J alarm values*

Service Water Effluent - - 4.B0E-04 µCi/cc 4 80E-06 µCi/cc

with effluent discharge not isolated (continued on next page)

PROCEDURE 5. 7 .1 REVISION 67 PAGE 17 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: The values listed for the Reactor Building vent, Turbine Building vent, and RW/ARW Building vent are based on a value of two times ODAM. The ERP UE value has been set to 1/5 the ODAM Instantaneous release limit to provide for a reasonable progression between the UE, Alert, and SAE classifications. If ERP UE value were placed at two times ODAM, the escalations between the UE, Alert, and SAE levels would have been significantly less than a decade apart. The ERP UE value allows approximately a 1 decade interval between the UE, Alert, and SAE classification points. Releases in excess of two times the site ODAM (Reference 3) instantaneous limits that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact the release was not isolated within 60 minutes. CNS Basis Reference(s):

1. Instrument Operating Procedure 4.15, Elevated Release Point and Building Kaman Radiation Monitoring Systems.

2. COR00l-18-01, Radiation Monitoring.

3. Off-Site Dose Assessment Manual - ODAM - For Assessment of Gaseous and Liquid Effluents at Cooper Nuclear Station.

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 18 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will likely exceed the applicable time. This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls. The ODAM multiples are specified in AU 1.1 and AAl.1 only to distinguish between non-emergency conditions and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate. This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the EAL. This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared. , PROCEDURE 5.7.1 REVISION 67 PAGE 19 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than two times the ODAM limits for 60 minutes or longer EAL: AU1.2 Unusual Event Any valid liquid effluent monitor reading > Table A-1 column "UE" for c: 60 min. (NOTE 2) NOTE 2 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume the release duration has exceeded the applicable time_ if an_ on-going_ release _is detected_ and _the_release start _time _is unknown. ________ _ Table A-1 Effluent Monitor Classification Thresholds GE SAE ALERT UE Monitor for~ 15 mm. for~ 15 mm. for~ 15 min. for~ 60 min. ERP 3.50E+08 µCi/sec 3.50E+07 µCl/sec 2.80E+06 µCl/sec 2.24E+05 µCi/sec Rx Bldg Vent 3 50E+07 µCi/sec 3.50E+06 µCl/sec 5.45E+05 µCi/sec 8.48E+04 µCi/sec V) 0 w Turb Bldg Vent 3 50E+07 µCi/sec 3.50E+06 µCl/sec 5.62E+05 µCi/sec 9.02E+04 µCi/sec V)

<(

(!) RW / ARW Bldg Vent 3.50E+07 µCl/sec 3.50E+06 µCi/sec 5 64E+05 µCi/sec 9.08E+04 µCi/sec C 200 x calculated a Rad Waste Effluent - - alarm values* 2 x calculated

J alarm values*

Service Water Effluent - - 4.B0E-04 µCi/cc 4.B0E-06 µCi/cc

with effluent discharge not isolated (continued on next page)

PROCEDURE 5.7.1 REVISION 67 PAGE 20 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: Liquid releases in excess of two times the site ODAM (Reference 3) instantaneous limits that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. CNS Basis Reference(s):

1. Instrument Operating Procedure 4.15, Elevated Release Point and Building Kaman Radiation Monitoring Systems.

2. Chemistry Procedure 8.8.11, Liquid Radioactive Waste Discharge Authorization.

3. COR00l-18-01, Radiation Monitoring.

4. Off-Site Dose Assessment Manual - ODAM - For Assessment of Gaseous and Liquid Effluents at Cooper Nuclear Station.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 21 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will likely exceed the applicable time. This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls. The ODAM multiples are specified in AU1.2 and AA1.2 only to distinguish between non-emergency conditions and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate. This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the EAL. This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared. PROCEDURE 5. 7 .1 REVISION 67 PAGE 22 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than two times the ODAM limits for 60 minutes or longer EAL: AU1.3 Unusual Event Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 x ODAM limits for~ 60 min. (NOTE 2) NOTE 2 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume the release duration has exceeded the applicable time_ if an_ on-going_ release _is detected_ and _the_ release start _time_ is unknown. ________ _ Mode Applicability: All CNS Basis: Releases in excess of two times the site Off-Site Dose Assessment Manual (ODAM) (Reference 1) instantaneous limits that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact the release was not isolated within 60 minutes. CNS Basis Reference(s):

1. Off-Site Dose Assessment Manual - ODAM - For Assessment of Gaseous and Liquid Effluents at Cooper Nuclear Station.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 23 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will likely exceed the applicable time. This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls. The ODAM multiples are specified in AU 1.3 and AAl.3 only to distinguish between non-emergency conditions and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate. This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). PROCEDURE 5.7.1 REVISION 67 PAGE 24 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODAM limits for 15 minutes or longer EAL: AA1.1 Alert Any valid gaseous monitor reading > Table A-1 column "Alert" for~ 15 min. (NOTE 2) NOTE 2 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume the release duration has exceeded the applicable _time_ if an_ on-going_ release _is detected_ and _the_ release start _time_ is unknown. _________ _ Table A-1 Effluent Monitor Classification Thresholds GE SAE ALERT UE Monitor for;:=,: 15 min. for~15min. for~ 15 min. for~ 60 min. ERP 3.50E+08 µCi/sec 3.50E+07 µCi/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec Rx Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.45E+05 µCi/sec 8.48E+04 µCi/sec en

J 0

w Turb Bldg Vent 3.50E+07 µCi/sec 3 50E+06 µCi/sec 5.62E+05 µO/sec 9.02E+04 µCi/sec en

<C

(!) RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec C 200 x calculated

J 0

Rad Waste Effluent - - alarm values* 2 x calculated

i alarm values*

Service Water Effluent - - 4.BOE-04 µCi/cc 4.B0E-06 µCi/cc

                                                                               * 'Mth effluent discharge not isolated (continued on next page)

PROCEDURE 5.7.1 REVISION 67 PAGE 25 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: If Alert thresholds were set at two-hundred times the ODAM limit, the resultant value would exceed the SAE threshold, which is based on 10% of the PAG limits. For this reason, the Alert gaseous thresholds have been set at the log-average of the UE threshold (two times ODAM limit for Turbine Building vent, Radwaste Building vent, and Reactor Building vent, 1/5 the ODAM limit for the ERP) and the

SAE threshold (CNS-DOSE dose assessment calculation) (Reference 1). This provides a reasonable escalation in classification from the UE to the Alert and SAE thresholds.

CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. Instrument Operating Procedure 4.15, Elevated Release Point and Building Kaman Radiation Monitoring Systems.

3. COR00l-18-01, Radiation Monitoring.

4. Off-Site Dose Assessment Manual - ODAM - For Assessment of Gaseous and Liquid Effluents at Cooper Nuclear Station.

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 26 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will likely exceed the applicable time. This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls. The ODAM multiples are specified in AU 1. 1 and Ml.1 only to distinguish between non-emergency conditions and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate. This EAL includes any release for which a radioactivity discharge permit was not prepared. This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared. This EAL directly correlates with the IC since annual average meteorology is required to be used in showing compliance with the ODAM and is used in calculating the Table A-1 setpoints. The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release. PROCEDURE 5. 7 .1 REVISION 67 PAGE 27 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODAM limits for 15 minutes or longer EAL: AAl.2 Alert Any valid liquid effluent monitor reading > Table A-1 column "Alert" for~ 15 min. (NOTE 2) NOTE 2- The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the release duration has exceeded or will likely exceed, the applicable time. In the absence of data to the contrary, assume the release duration has exceeded the applicable time_ if an_ on-going_ release _is detected_ and _the_ release start _time _Is unknown. __________: Table A-1 Effluent Monitor Classlflcatlon Thresholds GE SAE ALERT UE Monitor for~ 15 min. for;:,: 15 min. for;:,: 15 mm for;:,: 60 min. ERP 3.50E+08 µa/sec 3.50E+07 µCl/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec Rx Bldg Vent 3.50E+07 µa/sec 3.50E+06 µCl/sec 5.45E+05 µCl/sec 8.48E+04 µCi/sec U) 0 w Turb Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.62E+05 µCi/sec 9.02E+04 µa/sec i RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec C 200 x calculated

, Rad Waste Effluent - - alarm values* 2 x calculated a alarm values*

i Service Water Effluent - - 4.80E-04 µCl/cc 4.80E-06 µCi/cc

with effluent dJscha rge not 1Solated (continued on next page)

PROCEDURE 5. 7 .1 REVISION 67 PAGE 28 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100. CNS Basis Reference(s):

1. Instrument Operating Procedure 4.15, Elevated Release Point and Building Kaman Radiation Monitoring Systems.

2. Chemistry Procedure 8.8.11, Liquid Radioactive Waste Discharge Authorization.

3. COR00l-18-01, Radiation Monitoring.

4. Off-Site Dose Assessment Manual - ODAM - For Assessment of Gaseous and Liquid Effluents at Cooper Nuclear Station.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 29 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will likely exceed the applicable time. This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls. The ODAM multiples are specified in AUl.2 and AAl.2 only to distinguish between non-emergency conditions and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate. This EAL includes any release for which a radioactivity discharge permit was not prepared or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. This EAL addresses radioactivity releases, that for whatever reason, cause Radwaste Effluent radiation monitor readings to exceed the threshold identified in the IC established by the radioactivity discharge permit. The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release. PROCEDURE 5.7.1 REVISION 67 PAGE 30 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the ODAM limits for 15 minutes or longer EAL: Ml.3 Alert Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODAM limits for z 15 min. (NOTE 2) NOTE 2 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume the release duration has exceeded the applicable time_ if an_ on-going_ release _is detected_ and _the_ release start _time _is _unknown. ________ _ Mode Applicability: All CNS Basis: None CNS Basis Reference(s):

1. Off-Site Dose Assessment Manual - ODAM - For Assessment of Gaseous and Liquid Effluents at Cooper Nuclear Station.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 31 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will likely exceed the applicable time. This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.

The ODAM multiples are specified in AUl.3 and AAl.3 only to distinguish between non-emergency conditions and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL includes any release for which a radioactivity discharge permit was not prepared or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). PROCEDURE 5.7.1 REVISION 67 PAGE 32 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 0.1 Rem TEDE or 0.5 Rem thyroid CDE for the actual or projected duration of the release EAL: ASL 1 Site Area Emergency Any valid gaseous monitor reading > Table A-1 column "SAE" for~ 15 min. (NOTE 1) NOTE 1 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_a pp Iica bl e ti me. ______________________________________________________________________________ _ If dose as~essment results are available, declaration should be based on dose assessment instead of radiation monitor values (see EAL AS1.2). Do not delay declaration awaiting dose assessment results. Table A-1 Effluent Monitor Classification Thresholds GE SAE ALERT UE Monitor for~ 15 min. for~ 15 min. for~ 15 min. for~ 60 min. ERP 3.50E+08 µCl/sec 3.50E+07 µCl/sec 2.80E+06 µCi/sec 2.24E+05 µO/sec Rx Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.45E+05 µCl/sec 8.48E+04 µCi/sec en 0 w Turb Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.62E+05 µCi/sec 9.02E+04 µCl/sec en

RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCl/sec 9.08E+04 µCi/sec C 200 x calculated 5 Rad Waste Effluent - - alann values* 2 x calculated a alann values*

i Service Water Effluent - - 4.BOE-04 µCi/cc 4.BOE-06 µCi/cc

wrth effluent discharge not isolated (continued on next page)

PROCEDURE 5. 7.1 REvISION 67 PAGE 33 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: Table A-1 Site Area Emergency thresholds have been determined using CNS-DOSE dose projection calculations (Reference 1). The Site Area Emergency effluent monitor readings are 1 decade less than the General Emergency values. For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant (Reference 1). CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. CNS Drawing DWG.2.2 (P3-A-45).

NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate the classification is not warranted or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL. PROCEDURE 5. 7 .1 REVISION 67 PAGE 34 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 0.1 Rem TEDE or 0.5 Rem thyroid CDE for the actual or projected duration of the release EAL: ASl.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 0.1 Rem TEDE or

> 0.5 Rem thyroid CDE at or beyond the site boundary Mode Applicability:

All CNS Basis: The dose rate EALs are based on a Site Boundary dose rate of 0.1 Rem/hr TEDE or 0.5 Rem/hr CDE thyroid, whichever is more limiting. Actual meteorology is specifically identified since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible. For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant. CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. CNS-DOSE.

3. CNS Drawing DWG.2.2 (P3-A-45).

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 35 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate the classification is not warranted or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL. PROCEDURE 5.7.1 REVISION 67 PAGE 36 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 :.. Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 0.1 Rem TEDE or 0.5 Rem thyroid CDE for the actual or projected duration of the release EAL: AS1.3 Site Area Emergency Field survey indicates closed window dose rates > 0.1 Rem/hr that is expected to continue for~ 60 min. at or beyond the site boundary (NOTE 1) OR Field survey sample analysis indicates thyroid CDE > 0.5 Rem for 1 hr of inhalation at or beyond the site boundary NOTE 1 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_ a p pl ica b le time. ______________________________________________________________________________ _ If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values (see EAL AS1.2). Do not delay declaration awaiting dose assessment results. Mode Applicability: All CNS Basis: The 0.5 Rem integrated CDE thyroid dose was established in consideration of the 1: 5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid exposure. In establishing the field survey emergency action levels, a duration of 1 hour ls assumed. Therefore, the dose rate EALs are based on a Site Boundary dose rate of 0.1 Rem/hr TEDE or 0.5 Rem for 1 hour of inhalation CDE thyroid, whichever is more limiting. For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 37 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. CNS Drawing DWG.2.2 (P3-A-45).

NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. PROCEDURE 5. 7 .1 REVISION 67 PAGE 38 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 1 Rem TEDE or 5 Rem thyroid CDE for the actual or projected duration of the release using actual meteorology EAL: AGl. 1 General Emergency Any valid gaseous monitor reading > Table A-1 column "GE" for~ 15 min. (NOTE 1) NOTE 1 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_a p pl icable time. ______________________________________________________________________________ _ If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values (see EAL AGl.2). Do not delay declaration awaiting dose assessment results. Table A-1 Effluent Monitor Classification Thresholds GE SAE ALERT UE Monitor for:2'.15 min. for:2'.15 min for;::: 15 min. for;::: 60 min. ERP 3.50E+08 µCi/sec 3.50E+07 µCi/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec Rx Bldg Vent 3 50E+07 µCl/sec 3.50E+06 µCi/sec 5 45E+05 µCi/sec 8.48E+04 µCl/sec en

J 0

w Turb Bldg Vent 3 50E+07 µCi/sec 3.50E+06 µCi/sec 5.62E+05 µCi/sec 9.02E+04 µCl/sec en c:( (!) RW / ARm Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCl/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec C 200 x calculated

J a Rad Waste Effluent - - alarm values* 2 x calculated

i alarm values*

Service Water Effluent - - 4.80E-04 µCl/cc 4.80E-06 µCi/cc

wrth effluent discharge not Isolated (continued on next page)

PROCEDURE 5.7.1 REVISION 67 PAGE 39 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: The Table A-1 General Emergency thresholds have been determined using CNS-DOSE dose projection calculations (Reference 1). The General Emergency effluent monitor readings are one decade greater than the Site Area Emergency values. For the purposes of this EAL, the Site Boundary for CNS ls a 1 mile radius around the plant (Reference 2). CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. CNS Drawing DWG.2.2 (P3-A-45).

3. Computer Dose Projection (CNS DOSE).

NEI 99-01 BASIS: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate the classification is not warranted or may indicate that a higher classification Is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL. PROCEDURE 5.7.1 REVISION 67 PAGE 40 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 1 Rem TEDE or 5 Rem thyroid CDE for the actual or projected duration of the release using actual meteorology EAL: AGl.2 General Emergency Dose assessment using actual meteorology indicates doses > 1 Rem TEDE or

> 5 Rem thyroid CDE at or beyond the site boundary Mode Applicability:

All CNS Basis: The General Emergency values are based on the boundary dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 1 Rem TEDE or 5 Rem CDE thyroid for the actual or projected duration of the release. Actual meteorology Is specifically identified since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible. For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant. CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. CNS DOSE.

3. CNS Drawing DWG.2.2 (P3-A-45).

(continued on next page) PROCEDURE 5.7.1 REvISION 67 PAGE 41 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate the classification is not warranted or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If results of these dose assessments are available when the classification is made ( e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL. PROCEDURE 5. 7 .1 REVISION 67 PAGE 42 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Off-Site Rad Conditions Initiating Condition: Off-site dose resulting from an actual or imminent release of gaseous radioactivity greater than 1 Rem TEDE or 5 Rem thyroid CDE for the actual or projected duration of the release using actual meteorology EAL: AGl.3 General Emergency Field survey results indicate closed window dose rates > 1 Rem/hr expected to continue for~ 60 min. at or beyond the site boundary (NOTE 1) OR Analyses of field survey samples indicate thyroid CDE > 5 Rem for 1 hr of inhalation at or beyond the site boundary NOTE 1 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_a p pl ica b le ti me. ______________________________________________________________________________ _ If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values (see EAL AGl.2). Do not delay declaration awaiting dose assessment results. Mode Applicability: All CNS Basis: The 5 Rem integrated CDE thyroid dose was established in consideration of the 1: 5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid exposure. In establishing the dose rate emergency action levels, a duration of 1 hour is assumed. Therefore, the dose rate EALs are based on a Site Boundary dose rate of 1 Rem/hr TEDE or 5 Rem for 1 hour of inhalation CDE thyroid, whichever is more limiting. For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 43 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. CNS Drawing DWG.2.2 (P3-A-45).

NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. r PROCEDURE 5. 7.1 REVISION 67 PAGE 44 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - On-Site Rad Conditions And Spent Fuel Pool Events Initiating Condition: Unplanned rise in plant radiation levels EAL: AU2.1 Unusual Event Unplanned water level drop in the reactor cavity or spent fuel pool as indicated by any of the following:

LI-86 (calibrated to 1001' elev.)
Spent fuel pool low level alarm
Visual observation AND Valid Area radiation monitor reading rise on RMA-RA-1 or RMA-RA-2 Mode Applicability:

All (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 45 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: Loss of inventory from the reactor cavity or Spent Fuel Pool (SFP) may reduce water shielding above spent fuel and cause unexpected increases in plant radiation. Classification as an Unusual Event is warranted as a precursor to a more serious event. The major item of concern on loss of inventory from the reactor cavity and SFP is to maintain adequate water level for personnel shielding and cooling of the irradiated fuel. Normal SFP water level is 37' 6 1/2" above the bottom. A low SFP level alarm can be determined by Annunciator 9-4-2/A-3, FUEL POOL COOLING TROUBLE, alarming due to Annunciator Panel 25-15, Fuel Pool Low Level at 4" below normal. Decreases in SFP water level can also be detected through visual observation. The Skimmer Surge Tank low level alarm (Annunciator 9-4-2/C-3 at 100 ft 3 in the skimmer surge tank, elevation 981' 3") alone may not be conclusive evidence of an uncontrolled loss of inventory from the SFP. SFP weir wall design should prevent inadvertent draining of the SFP through Fuel Pool Cooling and Demineralizer System connections. A Skimmer Surge Tank low level alarm needs to be confirmed by visual observation to determine the extent of inventory loss from the SFP (Reference 1, 3, 4). During refueling when the RPV head is removed, Shutdown Range RPV water level instrument NBI-LI-86 is recalibrated to read vessel cavity level up to the 1001' elevation (Refuel Floor). With reactor cavity in communication with the Spent Fuel Pool via the fuel transfer canal, uncontrolled inventory loss can be remotely monitored via this indicator. NBI-LI-86 can be used only if it has been set up to read to 1001' elevation as specified in Procedure 4.6.1, Reactor Vessel Water Level Indication (Reference 5). Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment. Technical Specification Section LCO 3. 7.6 (Reference 7) requires spent fuel storage pool water level be maintained at least 21' 6" over the top of Irradiated fuel assemblies seated in the spent fuel storage pool racks. Technical Specification LCO 3.9.6 (Reference 8) requires RPV water level to be maintained at least 21' above the top of the RPV flange. During refueling, this maintains sufficient water level in the refueling cavity and SFP to retain iodine fission product activity in the water in the event of a fuel handling accident. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 46 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Area radiation monitors that may indicate a loss of shielding of spent fuel in the SFP or refueling cavity include (Reference 2, 6):

RMA-RA-1 (1448) RX BLDG FUEL POOL (HR) AREA.
RMA-RA-2 (1449) RX BLDG FUEL POOL (LR) AREA.

Portable radiation monitors are routinely employed to conduct radiation surveys in the Reactor Building. This source of information should not be excluded when considering emergency classification under this EAL, particularly when RMA-RA-1 and RMS-RA-2 may be taken out of service for preventative or corrective maintenance. CNS Basis Reference(s):

1. System Operating Procedure 2.2.32, Fuel Pool Cooling and Demineralizer System.

2. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1.

3. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2, C-3.

4. Abnormal Procedure 2.4FPC, Fuel Pool Cooling Trouble.

5. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

6. Emergency Procedure 5.lRAD, Building Radiation Trouble.

7. Technical Specification LCO 3. 7 .6.

8. Technical Specification LCO 3.9.6.

NEI 99-01 Basis: This EAL addresses increased radiation levels as a result of water level decreases above irradiated fuel. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant. The refueling pathway is the combination of refueling cavity and spent fuel pool. While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 4 7 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] For example, a refueling bridge ARM reading may increase due to planned evolutions such as head lift or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact be properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head. Generally, increased radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss. For refueling events where the water level drops below the RPV flange classification would be via CU2.1. This event escalates to an Alert via AA2.1 if irradiated fuel outside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be per the Fission Product Barrier Table for events in operating Modes 1-3. PROCEDURE 5. 7 .1 REVISION 67 PAGE 48 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - On-Site Rad Conditions And Spent Fuel Pool Events Initiating Condition: Unplanned rise in plant radiation levels EAL: AU2.2 Unusual Event Unplanned Valid Area radiation monitor reading or survey results rise by a factor of 1,000 over normal levels*

Normal levels can be considered as the highest reading in the past 24 hours excluding the current peak value Mode Applicability:

All CNS Basis: It is recognized that some plant area radiation monitors may not be able to detect or display a reading that is one-thousand times NORMAL LEVELS. The intent of this IC is to rely on currently installed plant monitors and not to require design changes/back fits. In cases where an installed area radiation monitor cannot detect or display values at or above 1,000 X NORMAL LEVELS value, then survey instrument results may be used. The ARMs monitor the gamma radiation levels in units of mR/hr at selected areas throughout the station. If radiation levels exceed a preset limit in any channel, the Control Room annunciator and local alarms will be energized to warn of abnormal or significantly changing radiological conditions. The alarm limit is normally set at approximately ten times normal background for each channel (Reference 1, 2). Routine and work specific surveys are conducted throughout the station at frequencies specified by RP Management. Routine surveys are scheduled per the RP Department surveillance schedule. Work specific surveys are conducted per the Radiation Work Permit (RWP) (Reference 3). (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 49 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s}:

1. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1.

2. Emergency Procedure 5. lRAD, Building Radiation Trouble.

3. Rad Protection Procedure 9.ALARA.4, Radiation Work Permits.

NEI 99-01 Basis: This EAL addresses increased radiation levels as a result of water level decreases above irradiated fuel or events that have resulted, or may result, in UNPLANNED increases in radiation dose rates within plant buildings. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant. This EAL addresses increases in plant radiation levels that represent a loss of control of radioactive material resulting in a potential degradation in the level of safety of the plant. This EAL excludes radiation level increases that result from planned activities such as use of radiographic sources and movement of radioactive materials. A specific list of ARMs is not required as it would restrict the applicability of the Threshold. The intent is to identify loss of control of radioactive material in any monitored area. PROCEDURE 5. 7 .1 REVISION 67 PAGE 50 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - On-Site Rad Conditions And Spent Fuel Pool Events Initiating Condition: Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the RPV EAL: AA2.1 Alert Damage to irradiated fuel OR loss of water level (uncovering irradiated fuel outside the RPV) that causes EITHER of the following: Valid RMA-RA-1 Fuel Pool Area Rad reading > 5.0E+04 mR/hr OR Valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum Hi-Hi alarm Mode Applicability: All CNS Basis: When considering classification, information may come from:

Radiation monitor readings.
Sampling and surveys.
Dose projections/calculations.
Reports from the scene regarding the extent of damage (e.g., Refueling Crew, Radiation Protection Technicians).

This EAL is defined by the specific areas where irradiated fuel is located, such as the refueling cavity or Spent Fuel Pool (SFP). The bases for the ventilation radiation Hi-Hi alarm is a spent fuel handling accident (Reference 2). Fuel Pool area radiation > 5.0E+04 mR/hr represents 100 times the high alarm setpoint (HR) and is unambiguously indicative of spent fuel damage or uncovery (Reference 1). (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 51 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1, Alarm A-10.

2. Alarm Procedure 2.3_9-4-1, Panel 9 Annunciator 9-4-1, Alarm E-4.

NEI 99-01 Basis: This EAL addresses increases in radiation dose rates within plant buildings and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant. This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage. Increased ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Increased background at the ventilation monitor due to water level decrease may mask increased ventilation exhaust airborne activity and needs to be considered. While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. Escalation of this emergency classification level, if appropriate, would be based on EALs in Subcategory Al. PROCEDURE 5.7.1 REVISION 67 PAGE 52 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - On-Site Rad Conditions And Spent Fuel Pool Events Initiating Condition: Damage to irradiated fuel or loss of water level that has or will result in the uncovering of Irradiated fuel outside the RPV EAL: AA2.2 Alert A water level drop in the reactor refueling cavity or spent fuel pool that will result in irradiated fuel becoming uncovered Mode Applicability: All CNS Basis: When considering classification, Information may come from:

Radiation monitor readings.
Sampling and surveys.
Dose projections/calculations.
Reports from the scene regarding the extent of damage (e.g., Refueling Crew, Radiation Protection Technicians).

The major item of concern on loss of inventory from the Spent Fuel Pool and Refueling Cavity is to maintain adequate water level for personnel shielding and cooling of the irradiated fuel. Normal Spent Fuel Pool water level is 37' 6 1/2" above the bottom. A low pool level alarm occurs at 4" below the normal water level. Decreases in Spent Fuel Pool water level can also be detected only through visual observation and the existence of the Skimmer Surge Tank low level alarm (9-4-2/C-3) at 100 ft 3 in the skimmer surge tank which is at elevation 981' 3" (Reference 1, 2, 3). (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 53 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] During refueling when the RPV head is removed, Shutdown Range RPV water level Instrument NBI-LI-86 is recalibrated to read vessel cavity level up to the 1001' elevation (Refuel Floor). With the reactor cavity in communication with the Spent Fuel Pool via the fuel transfer canal, uncontrolled inventory loss can be remotely monitored per this indicator. NBI-LI-86 can be used only if it has been set up to read to 1001' elevation as specified in Procedure 4.6.1, Reactor Vessel Water Level Indication (Reference 4). Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment. Technical Specification Section LCO 3. 7.6 (Reference 5) requires spent fuel storage pool water level be maintained at least 21' 6" over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. Technical Specification LCO 3.9.6 (Reference 6) requires RPV water level to be maintained at least 21' above the top of the RPV flange. During refueling, this maintains sufficient water level in the refueling cavity and SFP to retain iodine fission product activity in the water in the event of a fuel handling accident. CNS Basis Reference(s):

1. System Operating Procedure 2.2.32, Fuel Pool Cooling and Demineralizer System.

2. Annunciator Procedure 2.3_ 9-4-2, Panel 9 Annunciator 9-4-2.

3. Abnormal Procedure 2.4FPC, Fuel Pool Cooling Trouble.

4. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

5. Technical Specification LCO 3.7.6.

6. Technical Specification LCO 3.9.6.

NEI 99-01 Basis: This EAL addresses increases in radiation dose rates within plant buildings and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant. Escalation of this emergency classification level, if appropriate, would be based on EALs in Subcategory Al. PROCEDURE 5.7.1 REVISION 67 PAGE 54 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - On-Site Rad Conditions And Spent Fuel Pool Events Initiating Condition: Rise in radiation levels within the facility that Impedes operation of systems required to maintain plant safety functions EAL: AA3.1 Alert Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety functions: Main Control Room (RM-RA-20) OR CAS Mode Applicability: All CNS Basis: Areas that meet this threshold include the Main Control Room and the Central Alarm Station. The Central Alarm Station is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. There are no permanently installed CAS Area radiation monitors that may be used to assess this EAL threshold. Therefore, this portion of the EAL threshold must be assessed using local radiation survey (Reference 1, 2). CNS Basis Reference(s):

1. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1, B-10.

2. Emergency Procedure 5. lRAD, Building Radiation Trouble.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 55 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL addresses increased radiation levels that impact continued operation in areas requiring continuous occupancy to maintain safe operation or to perform a safe shutdown. The cause and/or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other EAL may be involved. The value of 15 mRem/hr is derived from the GDC 19 value of 5 Rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides the 15 mRem/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert. Areas requiring continuous occupancy include the Main Control Room and Central Alarm Station (CAS). PROCEDURE 5. 7 .1 REVISION 67 PAGE 56 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: AC power capability to critlcal buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in loss of all AC power to critical buses EAL: CU 1. 1 Unusual Event AC power capability to critical 4160V Buses lF and lG reduced to a single power source (Table C-4) for~ 15 min. such that any additional single failure would result in loss of all AC power to critical buses (NOTE 3) NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_a ppl ica ble ti me. ______________________________________________________________________________ _ Table C-4 AC Power Sources Offsite

Startup Station Service Transformer
Emergency Station Service Transformer
Backfeed 345 kv line through Main Power Transformer to the Normal Station Service Transformer (Note 8)

Onsite

DG-1
DG-2 NOTE 8 - The time required to establish the backfeed is likely longer than the specified time interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site power_ source._____________________________________________________________________________________________________________ _

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 57 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: The 4160V Critical Buses 1F (Div 1) and 1G (Div 2) are the plant essential, safety-related emergency buses. Each can be energized manually and separately by any of the following off-site sources of power: Figure C-1 illustrates the 4160V AC Distribution System (Reference 7, 8).

Startup Transformer - The Startup Transformer provides a source of off-site AC power to the entire Auxiliary Power Distribution System adequate for the startup operation or shutdown operation of the station. The Startup Transformer is the preferred source of off-site AC power to the station whenever the main generator is off-line. The Startup Transformer is energized from the 161 kV Switchyard. The transformer is normally left energized at all times to provide for quick automatic transfer of the 4160V auxiliaries to the Startup Transformer in the event the station Normal Transformer fails or the main generator trips off-line.
Emergency Transformer - The Emergency Transformer is the primary off-site AC power source to essential station loads. During normal station operation, the Emergency Transformer is energized by the 69 kV transmission line from OPPD.

As such, It supplies 4160V Switchgear 1F and/or 1G in the event the Normal Transformer and Startup Transformer are not available for service. Use of the Emergency Transformer also allows portions of the 345 kV System to be removed from service for inspection, testing, and maintenance.

Backfeeding power from the 345 kV line through the Main Power Transformer to the Normal Transformer. The Normal Transformer is the normal source of AC power to the station when the Main Generator is on-line above 20% (160 MWe) electrical power. The transformer Is energized during Main Generator operation through the Isolated Phase Buses that feed the Main Power Transformers. As mentioned in NOTE 8, the time required to establish the backfeed is likely longer than the 15 minute interval. If off-normal plant conditions have already established the backfeed its power to the safety-related buses may be considered an off-site power source.

On-site power sources are the emergency diesel generators (DG-1 and DG-2). (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 58 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If critical bus AC power Is reduced to a single source for > 15 minutes, an Unusu,al Event is declared under this EAL. If a power supply should have picked up a critical bus but failed to do so, that power supply should be considered unavailable until It has been successfully tied onto the bus. If power supply can be tied to the bus within 15 minutes, then power supply Is to be considered available. The Supplemental Diesel Generator (SDG) is not considered an on-site or an off-site emergency power supply and Is not considered in classifications involving loss of power. The SBO Coping Time per Regulatory Guide 1.155 considers the impact of a SDG. This cold condition EAL is equivalent to the hot condition loss of AC power EAL SAl.1. CNS Basis Reference(s):

1. System Operating Procedure 2.2.15, Startup Transformer.

2. System Operating Procedure 2.2.16, Normal Station Service Transformer.

3. System Operating Procedure 2.2.17, Emergency Station Service Transformer.

4. System Operating Procedure 2.2.18.1, 4160V Auxiliary Power Distribution System.

5. System Operating Procedure 2.2.18.3.DIVl, 4160V Div 1 Distribution Support.

6. System Operating Procedure 2.2.18.3.DIV2, 4160V Div 2 Distribution Support.

7. System Operating Procedure 2.2.18.4, 4160V Distribution Abnormal Power.

8. System Operating Procedure 2.2.20, Standby AC Power System (Diesel Generator).

9. Emergency Procedure 5.3SBO, Station Blackout.

10. BR 3001, One Line Diagram.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 59 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

11. BR 3002, Sheet 1.

12. NC43456, One Line Switching Diagram 161kV Substation.

13. Enercon Services, Inc. Report No. NPPl-PR-01, Station Blackout Coping Assessment for Cooper Nuclear Station, Revision 2.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 60 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure C-1: 4160V AC Distribution System FROM FROM MAIN GENERATOR 345 KV/161 KV GRID y NORMAL STARTUP y

                                                                                           \AAA/

STATION SERVICE

          ~        STATION SERVICE                                       TRANSFORMER       rvvv'\

TRANSFORMER 161 KV/4160V 22 KV/4160V 4160V SWGR 1C 4160V SWGR 1D BKR 1AN 4160V SWGR 1A ')BKR1BG 4160V SWGR 18 4160V 6WGR 1E BKR 1FA 4160V SWGR 1F 4160V SWGR 1G BKR 1FE ' ) BKR') 1FS ')BKR1GS ')BKR 1GE I I

              ~)                        1 =~.:
                                      )¢¢¢(STATION SERVICE I

MOOS

BKR EG2 c5 c5 TRANSFORMER SDG DIESEL GENERATOR #1 DIESEL GENERATOR #2 (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 61 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The condition indicated by this EAL is the degradation of the off-site and on-site AC power systems such that any additional single failure would result In a loss of all AC power to critical buses. This condition could occur due to a loss of off-site power with a concurrent failure of all but one emergency generator to supply power to its emergency bus. The subsequent loss of this single power source would escalate the event to an Alert in accordance with CAl .1. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. PROCEDURE 5.7.1 RE\/ISION 67 PAGE 62 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: Loss of all off-site and all on-site AC power to critical buses for 15 minutes or longer EAL: CAl.1 Alert Loss of all off-site and all on-site AC power (Table C-4) to critical 4160V Buses lF and lG for~ 15 min. (NOTE 3) NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_ applicable ti me. ______________________________________________________________________________ _

                                  ,Table C-4 AC Power Sources e,

Offsite

Startup Station Service Transformer
Emergency Station Service Transformer
Backfeed 345 kv line through Main Power Transformer to the Normal Station Service Transformer (Note 8)

Onsite

DG-1 /
DG-2 NOTE 8- The time required to establish the backfeed is likely longer than the specified time interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site power_ source._____________________________________________________________________________________________________________ _

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 63 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, D - Defueled CNS Basis: The 4160V Critical Buses lF (Div 1) and lG (Div 2) are the plant essential, safety-related emergency buses. Each can be energized manually and separately by any of the following off-site sources of power: Figure C-1 illustrates the 4160V AC Distribution System (Reference 7, 8).

Startup Transformer - The Startup Transformer provides a source of off-site AC power to the entire Auxiliary Power Distribution System adequate for the startup operation or shutdown operation of the station. The Startup Transformer is the preferred source of off-site AC power to the station whenever the main generator is off-line ( < 160 MWe). The Startup Transformer is energized from the 161 kV Switchyard. The transformer is normally left energized at all times to provide for quick automatic transfer of the 4160V auxiliaries to the Startup Transformer in the event the station Normal Transformer fails or the main generator trips off-line.
Emergency Transformer - The Emergency Transformer is the primary off-site AC power source to essential station loads. During normal station operation, the Emergency Transformer is energized by the 69 kV transmission line from OPPD. As such, it supplies 4160V Switchgear lF and/or lG in the event the Normal Transformer and Startup Transformer are not available for service. Use of the Emergency Transformer also allows portions of the 345 kV System to be removed from service for inspection, testing, and maintenance.
Backfeeding power from the 345 kV line through the Main Power Transformer to the Normal Transformer. The Normal Transformer is the normal source of AC power to the station when the Main Generator is on line above 20%

(160 MWe) electrical power. The transformer is energized during Main Generator operation through the Isolated Phase Buses that feed the Main Power Transformers. As mentioned in NOTE 8, the time required to establish the backfeed is likely longer than the 15 minute interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site power source. On-site power sources are the emergency diesel generators (DG-1 and DG-2). (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 64 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If a power supply should have picked up a critical bus but failed to do so, that power supply should be considered unavailable until it has been successfully tied onto the bus. If power supply can be tied to the bus within 15 minutes, then power supply is to be considered available. The Supplemental Diesel Generator (SDG) is not considered an on-site or an off-site emergency power supply and is not considered in classifications involving loss of power. The SBO Coping Time per Regulatory Guide 1.155 considers the impact of a SDG. This EAL is the cold condition equivalent of the hot condition loss of all AC power EAL 551.1. When in Cold Shutdown, Refueling, or Defueled Mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency buses, relative to that existing when in hot conditions. CNS Basis Reference(s):

1. System Operating Procedure 2.2.15, Startup Transformer.

2. System Operating Procedure 2.2.16, Normal Station Service Transformer.

3. System Operating Procedure 2.2.17, Emergency Station Service Transformer.

4. System Operating Procedure 2.2.18.1, 4160V Auxiliary Power Distribution System.

5. System Operating Procedure 2.2.18.3.DIVl, 4160V Div 1 Distribution Support.

6. System Operating Procedure 2.2.18.3.DIV2, 4160V Div 2 Distribution Support.

7. System Operating Procedure 2.2.18.4, 4160V Distribution Abnormal Power.

8. System Operating Procedure 2.2.20, Standby AC Power System (Diesel Generator).

9. Emergency Procedure 5.3SBO, Station Blackout.

10. BR 3001, One Line Diagram.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 65 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

11. BR 3002, Sheet 1.

12. NC43456, One Line Switching Diagram 161kV Substation.

13. Enercon Services, Inc. Report NPPl-PR-01, Station Blackout Coping Assessment for Cooper Nuclear Station, Revision 2.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 66 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure C-1: 4160V AC Distribution System FROM FROM MAIN GENERATOR 345 KV/161 KV GRID y NO~ STARTUP y v..A.AJ STATION SERVICE

          ~        STATION SERVICE                                      TRANSFORMER       rYVV'\

TRANSFORMER 161 KV/4160V 22 KV/4160V BKR 1CN

                         )
                                                               )~                 )~

i 4160V SWGR 1C ~ 4160VSWGR 1D ~ BKR 1AN 4160V SWOR 1A ')BKR ')BKR 4160VSWGR 1B 1BE 1BG 4160V SWGR 1E BKR') 1FA 4160VSWGR 1F

                                 ~                                                     4160VSWGR 1G 1FE ' )

BKR BKR') 1FS ')BKR1GS

                                                                                        ')BKR 1GE I                                 I I~~~

MODS BKR EG1

                     )
                                      ~STATION SERVICE                                    )~

I c5 c5 TRANSFORMER SOG DIESEL GENERATOR #1 DIESEL GENERATOR #2 (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 67 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, containment heat removal, spent fuel heat removal, and the ultimate heat sink. The event can be classified as an Alert when In Cold Shutdown, Refueling, or Defueled Mode because of the significantly reduced decay heat and lower temperature and pressure, increasing the time to restore one of the critical buses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels/Radiological Effluent EALs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. PROCEDURE 5. 7 .1 REVISION 67 PAGE 68 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL: CU2.1 Unusual Event RPV level cannot be restored and maintained > +3 in. for~ 15 min. (NOTE 3) due to RCS leakage NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii ke ly exceed the_ a pp Ii cable time. ______________________________________________________________________________ _ Mode Applicability: 4 - Cold Shutdown CNS Basis: The condition of this EAL may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. When RPV level drops to +3 inches (low level scram setpoint), level is well below the normal control band and automatic RPS and PCIS actuations are required (Reference 3, 5). RPV level is normally monitored using the following instruments (Reference 1, 2):

Wide Range NBI-Ll-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-Ll-92 (0 to 180 inches).
Fuel Zone Range NBI-Ll-91A, B & C (-320 to 60 inches).
Narrow Range RFC-Ll-94A, B & C (0 to 60 inches).
Shutdown Range NBI-Ll-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 69 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] This Cold Shutdown EAL represents the hot condition EAL SU6.1, in which RCS leakage is associated with Technical Specification limits. In Cold Shutdown, these limits are not applicable; hence, the use of RPV level as the parameter of concern in this EAL (Reference 4). CNS Basis Reference(s):

1. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

2. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

3. EOP-1A RPV Control.

4. NEI/NRC EAL FAQ #2006-014.

5. Technical Specification Table 3.3.1.1-1.

NEI 99-01 Basis: This EAL is considered to be a potential degradation of the level of safety of the plant. The inability to maintain or restore level is indicative of loss of RCS inventory. Relief valve normal operation should be excluded from this EAL. However, a relief valve that operates and fails to close per design should be considered applicable to this EAL if relief valve cannot be isolated. Prolonged loss of RCS inventory may result in escalation to the Alert emergency classification level via either CA2.1 or CA3. l. The difference between CU2.1 and CU2.2 deals with the RCS conditions that exist between Cold Shutdown and Refueling Modes. In Cold Shutdown, the RPV will normally be intact and RPV level is typically controlled below the elevation of the RPV flange and above the low-end of the normal control band. In the Refueling Mode the RPV is not intact and any planned evolutions to lower RPV level below the elevation of the RPV flange must be carefully controlled. PROCEDURE 5.7.1 REVISION 67 PAGE 70 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL: CU2.2 Unusual Event Unplanned RPV level drop for~ 15 min. (NOTE 3) below EITHER: RPV flange (LI-86: 206 in. normal calibration, 113. 75 in. elevated calibration) OR RPV level band when the RPV level band is established below the RPV flange NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it Is determined the condition will Ii kely exceed the_ a p pl ica ble time. ______________________________________________________________________________ _ Mode Applicability: 5 - Refueling CNS Basis: The RPV flange is 722.75 Inches above the RPV bottom head. RPV water level at this elevation is normally indicated by the Shutdown Range instrument (LI-86 Shutdown, 0 - 400 inches). When calibrated for normal plant operations, the Shutdown Range instrument reads 206 inches at the RPV flange. With the RPV head removed, the instrument is calibrated to indicate reactor cavity water levels as high as the refuel floor. When calibrated for elevated indication, the Shutdown Range instrument reads 113. 75 inches at the RPV flange. Visual observation of water level in the reactor cavity and RPV is also used during refuel operations. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 71 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. General Operating Procedure 2.1.20.3, RPV Refueling Preparation (Wet Lift or Dryer and Separator), Indicated RPV Level vs. RPV Height attachment.

2. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

3. IAC Procedure 14.15.3, Reactor Vessel Open Head Level Monitor System.

NEI 99-01 Basis: This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RPV water level below the RPV flange are carefully planned and procedurally controlled. An unplanned event that results in water level decreasing below the RPV flange or below the planned RPV water level for the given evolution (if the planned RPV water level is already below the RPV flange) warrants declaration of an Unusual Event due to the reduced RCS inventory that is available to keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame, it may indicate a more serious condition exists. Continued loss of RCS inventory will result in escalation to the Alert emergency classification level via either CA2.1 or CA3. l. The difference between CU2.1 and CU2.2 deals with the RPV conditions that exist between Cold Shutdown and Refuel Modes. In Cold Shutdown, the RCS will normally be intact and standard RPV inventory and level monitoring means are available. In Refuel Mode, the RCS is not intact and RPV level and inventory may be monitored by different means. This EAL involves a decrease in RPV level below the top of the RPV flange or a decrease below the RPV level band (when the RPV level band Is established below the RPV flange) that continues for 15 minutes due to an unplanned event. This EAL is not applicable to decreases in flooded reactor cavity level, which is addressed by AU2.1, until such time as the level decreases to the level of the vessel flange. , If RPV level continues to decrease and reaches the Low-Low ECCS actuation setpoint, then escalation to CA2.1 would be appropriate. PROCEDURE 5. 7 .1 REVISION 67 PAGE 72 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

Category: C - Cold Shutdown/Refueling System Malfunctlon Subcategory: 2 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL: CU2.3 Unusual Event RPV level cannot be monitored with any unexplained RPV leakage indication, Table C-1 Tab_le C-1 RPV Leakage Indications

Drywall equipment drain sump level rise
Drywell floor drain sump level rise
Reactor Building equipment drain sump level rise
Reactor Building floor drain sump level rise
Torus water level rise
RPV make-up rate rise
Observation of unisolable RCS leakage Mode Applicability:

5 - Refueling CNS Basis: RPV level is normally monitored using the following instruments (Reference 3, 4):

Wide Range NBI-LI-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-LI-92 (0 to 180 inchesf
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 inches).
Narrow Range RFC-LI-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 73 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. In this EAL, all water level indication ls unavailable and the RPV inventory loss should be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (Reference 1). A Reactor Building equipment or floor drain sump level rise may also be Indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling Mode, an unexplained rise in torus water level could be indicative of RHR valve misalignment or leakage (Reference 2). If make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory. CNS Basis Reference(s):

1. System Operating Procedure 2.2.27, Equipment, Floor, and Chemical Drain System.

2. System Operating Procedure 2.2.69, Residual Heat Removal System.

3. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

4. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

NEI 99-01 Basis: This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant. Continued loss of RCS Inventory will result in escalation to the Alert emergency classification level via either CA2.1 or CA3.1. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 74 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] This EAL addresses conditions in the Refueling Mode when normal means of core temperature indication and RPV level Indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the Operators would need to determine that RPV inventory loss was occurring by observing sump level changes listed in Table C-1. Sump 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 RCS leakage. Escalation to the Alert emergency classification level would be via either CA2.1 or CA3.l. PROCEDURE 5. 7 .1 REVISION 67 PAGE 75 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory EAL: CA2.1 Alert RPV level < -42 in. OR RPV level cannot be monitored for~ 15 min. (NOTE 3) with any unexplained RPV leakage indication, Table C-1 NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_ a pp Ii cab Ie time. ______________________________________________________________________________ _ Table C-1 RPV Leakage Indications

Drywell equipment drain sump level rise
Drywell floor drain sump level rise
Reactor Building equipment drain sump level rise
Reactor Building floor drain sump level rise
Suppression pool water level rise
RPV make-up rate rise
Observation of unisolable RCS leakage Mode Applicability:

4 - Cold Shutdown, 5 - Refueling (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 76 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: The threshold RPV level of -42 inches is the low-low ECCS actuation setpoint (Reference 5). RPV level is normally monitored using the following instruments (Reference 3, 4):

Wide Range NBI-LI-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-LI-92 (0 to 180 inches).
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 inches).
Narrow Range RFC-LI-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (Reference 1). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling Mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (Reference 2). If make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 77 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. System Operating Procedure 2.2.27, Equipment, Floor, and Chemical Drain System.

2. System Operating Procedure 2.2.69, Residual Heat Removal System.

3. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

4. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

5. Technical Specification Table 3.3.5.1-1.

NEI 99-01 Basis: This EAL serves as precursor to a loss of heat removal. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum classification of Alert. The low-low ECCS Actuation setpoint was chosen because it Is a recognized setpoint. The inability to restore and maintain level after reaching this setpoint would therefore be indicative of a failure of the RCS barrier. In Cold Shutdown Mode, the RCS will normally be intact and standard RPV level monitoring means are available. In the Refueling Mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually. In the Refueling Mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the Operators would need to determine that RPV inventory loss was occurring by observing sump level changes listed in Table C-1. Sump level increas*es must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 15 minute duration for the loss of level indication was chosen because it is half of the CS2.1 Site Area Emergency EAL duration. The 15 minute duration allows CA2.1 to be an effective precursor to CS2.1. Significant fuel damage is not expected to occur until the core has been uncovered for > 1 hour. Therefore, this EAL meets the definition for an Alert. If RPV level continues to decrease, then escalation to Site Area Emergency will be via EALCS2.1. PROCEDURE 5.7.1 REVISION 67 PAGE 78 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV Inventory affecting core decay heat removal capability EAL: CS2.1 Site Area Emergency With Containment Closure not established (NOTE 4), RPV level < -48 in. NOTE 4- Containment Closure is the action taken to secure primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure_ 0. 50. 5, Outage_ Shutdown_ Safety.--------------------------------------------------------------- Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: When RPV level decreases to -48 inches, water level is 6 Inches below the low-low ECCS actuation setpoint: -42 inches - 6 inches = -48 inches (Reference 4). RPV level is normally monitored using the following instruments (Reference 2, 3):

Wide Range NBI-LI-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-LI-92 (0 to 180 inches).
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 inches).
Narrow Range RFC-LI-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 Indicates when an instrument may be used for level indication in the EOPs/SAGs. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 79 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier. Containment Closure is the action taken to secure Primary Containment or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure 0.50.5, Outage Shutdown Safety (Reference 1). CNS Basis Reference(s):

1. Administrative Procedure 0.50.5, Outage Shutdown Safety.

2. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

3. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

4. Technical Specification Table 3.3.5.1-1.

NEI 99-01 Basis: Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to a RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted. Escalation to a General Emergency is via CG2.1 or AGl.1. PROCEDURE 5. 7.1 REVISION 67 PAGE 80 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS2.2 Site Area Emergency With Containment Closure established (NOTE 4), RPV level < -158 in. NOTE 4- Containment Closure is the action taken to secure primary or Secondary Containment and Its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified In Administrative Procedure_ 0. 50. 5, Outage_ Shutdown_Safety.-------------------------------------------------------------- Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: When RPV level drops to -158 inches, core uncovery is about to occur (Reference 4). RPV level is normally monitored using the following instruments (Reference 2, 3):

Wide Range NBI-LI-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-LI-92 (0 to 180 inches).
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 Inches).
Narrow Range RFC-LI-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 81 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint Infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier. Containment Closure is the action taken to secure Primary Containment or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure 0.50.5, Outage Shutdown Safety (Reference 1). CNS Basis Reference(s):

1. Administrative Procedure 0.50.5, Outage Shutdown Safety.

2. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

3. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

4. NEDC 97-089.

NEI 99-01 Basis: Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to a RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted. Escalation to a General Emergency is via CG2. lor AGl. 1. PROCEDURE 5.7.1 REVISION 67 PAGE 82 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS2.3 Site Area Emergency RPV level cannot be monitored for~ 30 min. (NOTE 3) with a loss of inventory as indicated by EITHER: Unexplained RPV leakage indication, Table C-1 OR Erratic Source Range Monitor indication NOTE 3- The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_ a p pli cable ti me. ______________________________________________________________________________ _ Table C-1 RPV Leakage Indications

                      * -Drywell equipment drain sump level rise

Drywell floor drain sump level rise
Reactor Building equipment drain sump level rise
Reactor Building floor drain sump level rise
Suppression pool water level rise
RPV make-up rate rise
Observation of unisolable RCS leakage (continued on next page)

PROCEDURE 5. 7 .1 REVISION 67 PAGE 83 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: RPV level is normally monitored using the instruments in Figure C-3 (Reference 3, 5).

Wide Range NBI-Li-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-U-92 (0 to 180 inches).
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 inches).
Narrow Range RFC-U-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1 or erratic Source Range Monitor (SRM) indication:

Table C-1 level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (Reference 1). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling Mode, an unexplained rise in torus water level could be indicative of RHR valve misalignment or leakage (Reference 2). If make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 84 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

Source Range Monitor (SRM) indication is provided in the Main Control Room by NMS-I-43A-D, SRM A-D LOG COUNT RATE, NM-NR-45, SRM 2 PEN RECORDER, and SPDS (Reference 4). Erratic Source Range Monitor Indication specified Is caused by increased boiling in the vicinity of the SRMs as water level lowers and exposed fuel heats up. This instability will be greatly magnified when compared with normal source range monitor behavior. Additionally, once the water level falls below the SRM level, lack of thermallzed neutron populations In the area will cause the SRMs to read abnormally low.

At some plants, containment high range area radiation monitors are specified as an alternate means of detecting loss of water shielding above the core and possible core uncovery. The CNS Primary Containment radiation monitors are designated RA-RM-40A and B. The Containment High Range Monitoring System monitors gamma radiation levels of 1 R/hr to 10 7 R/hr in the reactor drywell area. The system consists of two ion chambers in the drywell and two readout modules in Panels 9-2 and 9-10. The detectors (RA-RE-40A and B) are located 180° from each other in the drywell above elevation 901' 6". The elevation of the bottom of the RPV bottom head is 917'. The elevation of the top of active fuel in the Primary Containment is~ 947'. The detectors are~ 45' 6" below the top of active fuel. RCS piping, components, and drywell structural members are positioned between the detectors and the reactor core. Due to the relative location of these detectors with respect to the top of active fuel, the CNS containment high range radiation monitors cannot be utilized for detection of loss of RPV inventory above the core. Additionally, no other installed Radiation Monitoring System exists that can be utilized for the function. CNS Basis Reference(s):

1. System Operating Procedure 2.2.27, Equipment, Floor, and Chemical Drain System.

2. System Operating Procedure 2.2.69, Residual Heat Removal System.

3. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

4. Instrument Operating Procedure 4.1.1, Source Range Monitoring System.

5. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 85 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to a RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted. Escalation to a General Emergency is via CG2.2 or AGl. 1. In the Cold Shutdown Mode, normal RPV Level Instrumentation Systems will usually be available. In the Refueling Mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the Operators would need to determine that RPV inventory loss was occurring by observing sump level changes listed in Table C-1. Sump 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 RCS leakage. The 30 minute duration allows sufficient time for actions to be performed to recover inventory control equipment. Post-TMI studies indicated the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. PROCEDURE 5. 7 .1 REVISION 67 PAGE 86 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL: CG2.1 General Emergency RPV level < -158 in. for~ 30 min. (NOTE 3) AND Any Containment Challenge indication, Table C-5 NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii ke ly exceed the_a pp Ii cab le ti me. ______________________________________________________________________________ _

                         - Table C-5 Con~inment Challenge Indications

Containment Closure not established (Note 4)
Deflagration concentrations exist inside PC

                               ~  6% H2 in drywall or torus AND
                               ~  5% 0 2 in drywell or torus

Unplanned rise in PC pressure
Secondary Containment area radiation

                            > 1000 mR/hr (EOP-5A Table 10)

NOTE 4- Containment Closure is the action taken to secure primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative i_ Procedure_ 0. 50. 5, Outage_ Shutdown_ Safety.------------------------------------------------------------ __ _ (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 87 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: When RPV level drops to -158 inches, core uncovery is about to occur (Reference 7). RPV level is normally monitored using the following instruments (Reference 3, 4):

Wide Range NBI-LI-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-LI-92 (0 to 180 inches).
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 inches).
Narrow Range RFC-LI-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. Four conditions are associated with a challenge to Primary Containment (PC) integrity:

1. Containment Closure is the action taken to secure Primary Containment or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure 0.50.5, Outage Shutdown Safety (Reference 1).

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 88 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

2. Deflagratlon (explosive) mixtures In the Primary Containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radlolysis ls a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction Is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to Primary Containment integrity. The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen) (Reference 5).

3. Any unplanned increase in PC pressure in the Cold Shutdown or Refueling Mode indicates a potential loss of Containment Closure capability. Unplanned Primary Containment pressure increases indicates Containment Closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

4. 1,000 mR/hr Is the Secondary Containment Maximum Safe Operating radiation value. Exceeding this value is indicative of problems in the Secondary Containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-SA, Secondary Containment Control, Table 10. As indicated by NOTE 5 in EOP-SA, Table 10, RP Surveys and ARM Teledoslmetry System may be used for these indications (Reference 6).

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 89 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] EOP-SA Table 10 - Secondary Containment Radiation Levels 1oI SECQ'iDARf CO\'TM'.'MCNT RArxATICN LcVE!..S S?DS 15 0 Muimun S:tfci W~-Ull'Jrn Na-ma Op:!rufug Val.JC! Op:!,.:ifug V.:ii.J<!

       ~                             ~:J~* !e\1     ~ra:tll~ f   Eh:i;i~ ¢:ll8!1*)  A:e.:i. ll i"I~ ¢:DB !:J""\

1

                                                                                                                 ~dtt1' 'l*1r.. 1::!

FUD.. Poet. /\ REA RMA,RA-1 6 1001' EL 1000 100

10 FU!:!. ?OCf_ AREA RMA*RA*2 .01
100 1001' El.

R','ICU ?RECOAT ARM RMA*RA~ 0.1

1000 gsa* EI.

RWCU S..UDGE A'{D DEC".NT PUM.PAREA R..\1-o\ ,RA--5 0.1

1000 9J1*a 1000 CRD HYDRAULIC EOU1?

AREA {SOUT!fi RMA-RAo ..01

100 903' 0.

CRD HYDRAULIC EOU1P AREA~ORTH} RMA*RA-9 .01

100 H?CI ?UMP ROC~~ RMA*RA-10 .01-100 H.?CI Room RHR Pl.IM? ROCM,

{SOUTHWEST} RMA-RA*11 .01

100 S\'/OJ~d 1000 TORUS H?V AR.EA

{SOUTif,'lEST} RMA*RA-27 1.0

10000 SWToCJ:1.

RHR ?UM? ROC>.\t. {NORTitWElD R&A-RA-12 .01 -100 NWQwd 1000 RC)C.'CORE SPRAY ?U!-1..P ROOM. {N:l RTIH:AST} RMA*RA-13 .01

100 NEOJal 1000 CORE S?RAY Pl.IM? ROO\!.,

(SO l.ITHEAST} RMA*RA-1l. .01

100 SEOwd 1000 NOTES Area radiation levels can be monitored by RP surveys or ARM teledos1metry system (continued on next page)

PROCEDURE 5. 7.1 REVISION 67 PAGE 90 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Administrative Procedure 0.50.5, Outage Shutdown Safety.

2. System Operating Procedure 2.2.60.1, Containment H2/'O2 Monitoring System.

3. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

4. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

5. AMP-TBD00, Step PC/H.

6. EOP-SA, Secondary Containment Control, Table 10.

7. NEDC 97-089.

NEI 99-01 Basis: This EAL represents the inability to restore and maintain RPV level to above the top of active fuel with containment challenged. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level. Analysis indicates that core damage may occur within an hour following continued core uncovery; therefore, conservatively, 30 minutes was chosen. The General Emergency is declared on the occurrence of the loss or imminent loss of function of all three barriers. Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure. With the Primary Containment and Secondary Containment breached or challenged, then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a General Emergency. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 91 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Containment Closure is the action taken to secure containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure should not be confused with refueling containment integrity as defined in Technical Specifications. Site shutdown contingency plans typically provide for re-establishing Containment Closure following a loss of heat removal or RPV inventory functions. If closure is re-established prior to exceeding the temperature or level thresholds of the RCS barrier and Fuel Clad barrier EALs, escalation to General Emergency would not occur. If Containment Closure is re-established prior to exceeding the 30 minute core uncovery time limit, then escalation to GE would not occur. The use of Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Secondary Containment. The radiation monitor values are based on the EOP "maximum safe values" because these values are easily recognizable and have an emergency basis. 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 gasses in Primary Containment. However, Primary 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. PROCEDURE 5.7.1 REVISION 67 PAGE 92 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL: CG2.2 General Emergency RPV level cannot be monitored for~ 30 min. (NOTE 3) with core uncovery indicated by EITHER: Unexplained RPV leakage indication, Table C-1 OR Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table C-5 NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_ a pp Ii cable time. ______________________________________________________________________________ _ Table C-1 RPV Leakage Indications

Drywell equipment drain sump level rise
Drywell floor drain sump level rise
Reactor Building equipment drain sump level rise
Reactor Building floor drain sump level rise
Suppression pool water level rise
RPV make-up rate rise
Observation of unisolable RCS leakage (continued on next page)

PROCEDURE 5.7.1 REVISION 67 PAGE 93 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Table C-5 Containment Challenge Indications

Containment Closure not established (Note 4)
Deflagration concentrations exist inside PC

                                    ~  6% H2 in drywell or torus AND
                                    ~  5% 0 2 in drywall or torus

Unplanned rise in PC pressure
Secondary Containment area radiation

                                 > 1000 mR/hr (EOP-5A Table 10)

NOTE 4- Containment Closure is the action taken to secure Primary or Secondary : Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure_ 0. 50. 5, Outage_ Shutdown_Safety.--------------------------------------------------------------- Mode Applicability:

4 - Cold Shutdown, 5 - Refueling CNS Basis:

RPV level is normally monitored using the following instruments (Reference 5, 7):

Wide Range NBI-LI-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-LI-92 (0 to 180 inches).
Fuel Zone Range NBI-LI-91A, B & C (-320 to 60 inches).
Narrow Range RFC-LI-94A, B & C (0 to 60 inches).
Shutdown Range NBI-LI-86 (0 to 400 inches).

Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs .. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 94 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1 or erratic Source Range Monitor (SRM) indication:

Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (Reference 2). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses 'external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling Mode, an unexplained rise in torus water level could be indicative of RHR valve misalignment or leakage (Reference 4). If make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be'isolated could be indicative of a loss of RPV inventory.

Source Range Monitor (SRM) indication is provided in the Main Control Room by NMS-I-43A-D, SRM A-D LOG COUNT RATE, NM-NR-45, SRM 2 PEN RECORDER, and SPDS (Reference 6). Erratic Source Range Monitor indication specified is caused by increased boiling in the vicinity of the SRMs as water level lowers and exposed fuel heats up. This instability will be greatly magnified when compared with normal source range monitor behavior. Additionally, once the water level falls below the SRM level, lack of thermalized neutron populations in the area will cause the SRMs to read abnormally low.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 95 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] At some plants, containment high range area radiation monitors are specified as an alternate means of detecting loss of water shielding above the core and possible core uncovery. The CNS Primary Containment radiation monitors are designated RA-RM-40A and RA-RM-40B. The Containment High Range Monitoring System monitors gamma radiation levels of 1 R/hr to 10 7 R/hr In the reactor drywell area. The system consists of two ion chambers in the drywell and two readout modules in Panels 9-2 and 9-10. The detectors (RA-RE-40A and RA-RE-40B) are located 180° from each other in the drywell above elevation 901' 6". The elevation of the bottom of the RPV bottom head is 917'. The elevation of the top of active fuel in the Primary Containment is ~ 947'. The detectors are ~ 45' 6" below the top of active fuel. RCS piping, components, and drywell structural members are positioned between the detectors and the reactor core. Due to the relative location of these detectors, with respect to the top of active fuel, the CNS containment high range radiation monitors cannot be utilized for detection of loss of RPV inventory above the core. Additionally, no other installed Radiation Monitoring System exists that can be utilized for the function. Four conditions are associated with a challenge to Primary Containment (PC) integrity:

1. Containment Closure is the action taken to secure Primary Containment or Secondary Containment and Its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure 0.50.5, Outage Shutdown Safety (Reference 1).

2. Deflagration ( explosive) mixtures in the Primary Containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to Primary Containment integrity. The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen) (Reference 8).

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 96 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

3. Any unplanned increase in PC pressure in the Cold Shutdown or Refueling Mode indicates a potential loss of Containment Closure capability. Unplanned Primary Containment pressure increases indicates Containment Closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

4. 1,000 mR/hr is the Secondary Containment Maximum Safe Operating radiation value. Exceeding this value is indicative of problems in the Secondary Containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-SA, Secondary Containment Control, Table 10. As indicated by NOTE 5 in EOP-SA, Table 10, RP Surveys and ARM Teledosimetry System may be used for these indications (Reference 9).

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 97 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] EOP-SA Table 10 - Secondary Containment Radiation Levels 10 SECONDARY CONTAINMENT RADIATION LEVELS SPDS15 0 Maximum Safe Maximum Normal Operating Value Operating Value Am.a 8 ll 8BM 81aanfld Baag1;: ,mBlbr) Ama 1!alu1;: (mBt'.bc) 81.tual 1!alu1;: FUEL POOL AREA RMA-RA-1 100-1D6 1001' El. 1000 FUEL POOL AREA RMA-RA-2 .01 - 100 1001' El. RWCU PRECOAT AREA RMA-RA-4 0.1 -1000 958' El. RWCU SLUDGE AND DECANT PUMP AREA RMA-RA-5 0.1 -1000 931' El. 1000 CRD HYDRAULIC EQUIP AREA (SOUTH) RMA-RA-8 .01 -100 903' El. CRD HYDRAULIC EQUIP AREA (NORTH) RMA-RA-9 .01 -100 HPCI PUMP ROOM RMA-RA-10 .01 -100 HPCI Room RHR PUMP ROOM, (SOUTHWEST) RMA-RA-11 .01 -100 SW Quad 1000 TORUS HPV AREA (SOUTHWESn RMA-RA-27 1.0 - 10000 SW Torus RHR PUMP ROOM, (NORTHWEST) RMA-RA-12 .01 -100 rwv Quad 1000 RCIC/CORE SPAAY PUMP ROOM, (NORTHEAST) RMA-RA-13 .01 -100 NE Quad 1000 CORE SPRAY PUMP ROOM, (SOUTHEAST) RMA-RA-14 .01 -100 SE Quad 1000 NOTES Area radiabon levels can be monitored by RP surveys or ARM teledos1metry system (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 98 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Administrative Procedure 0.50.5, Outage Shutdown Safety.

2. System Operating Procedure 2.2.27, Equipment, Floor, and Chemical Drain System.

*3, System Operating Procedure 2.2.60.1, Containment Hi/O2 Monitoring System.

4. System Operating Procedure 2.2.69, Residual Heat Removal System.

5. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

6. Instrument Operating Procedure 4.1.1, Source Range Monitoring System.

7. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

8. AMP-TBD00, Step PC/H.

9. EOP-5A, Secondary Containment Control, Table 10; NEI 99-01 Basis:

In the Cold Shutdown Mode, nqrmal RPV level and RPV Level Instrumentation Systems will normally be available. However, if all level indication were to be lost during a loss of RPV inventory event, the Operators would need to determine that RPV inventory loss was occurring by observing sump level changes. Sump level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the Primary Containment to ensure they are indicative of RCS leakage. In the Refueling Mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the Operators would need to determine that RPV inventory loss was occurring by observing sump level changes listed in Table C-1. For both Cold Shutdown and Refueling Modes, sump level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the Primary Containment to ensure they are indicative of RCS leakage. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 99 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] A number of variables, such as initial RPV level or shutdown heat removal system design, can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that significant core damage may occur within an hour following continued core uncovery; therefore, conservatively, 30 minutes was chosen. Post-TMI studies indicated the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. The General Emergency is declared on the occurrence of the loss or imminent loss of function of all three barriers. Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure. With the Primary Containment and Secondary Containment breached or challenged, then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a General Emergency. Containment Closure is the action taken to secure either Primary Containment or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure should not be confused with refueling containment integrity as defined in Technical Specifications. Site shutdown contingency plans provide for re-establishing Containment Closure following a loss of heat removal or RPV inventory functions. If closure is re-established prior to exceeding the temperature or level thresholds of the RCS Barrier and Fuel Clad Barrier EALs, escalation to General Emergency would not occur. If Containment Closure is re-established prior to exceeding the 30 minute core uncovery time limit, then escalation to GE would not occur. The use of Secondary Containment radiation monitors should provide Indication of increased release that may be Indicative of a challenge to Secondary Containment. The radiation monitor values are based on the EOP "maximum safe values" because these values are easily recognizable and have an emergency basis. 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 gasses in Primary Containment. However, Primary 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. PROCEDURE 5. 7.1 REVISION 67 PAGE 100 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL: CU3.1 Unusual Event Any unplanned event results in RCS temperature > 212°F due to loss of decay heat removal capability Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: Several instruments are capable of providing Indication of RPV temperature with respect to the Technical Specification cold shutdown temperature limit (212°F) (Reference 4). These include (Reference 1, 2):

NBI-TR-89, REACTOR VESSEL METAL TEMPERATURE RECORDER (Panel 9-21).
Vessel Drain, PMIS Point M180, or NBI-TR Point 06 if M180 is not available.
Vessel Bottom Head, PMIS Point M184, or NBI-TR Point 10 if M184 is not available.
Bottom Head Adjacent to Support Skirt, PMIS Point M183, or NBI-TR Point 09.
RR-TR-165, RR SUCTION & FEEDWATER TEMP (Panel 9-4).

PMIS Points M174 through M185 can be used to monitor RPV temperatures. Thermocouples associated with computer Points M180, M183, and M185 do not respond as quickly nor register as high a temperature as other thermocouples due to their locations. Inservice leak testing, hydrostatic testing, and control rod scram time testing in which RCS temperature is intentionally raised above 212°F per Technical Specification LCO 3.10.1 are not applicable to this EAL (Reference 3). (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 101 OF 342

ATTACHMENT 2 EMEzRGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. General Operating Procedure 2.1.1, Startup Procedure.

2. System Operating Procedure 2.2.69.2, RHR System Shutdown Operations.

3. Technical Specifications LCO 3.10.1.

4. Technical Specifications Table 1.1-1.

NEI 99-01 Basis: This EAL is an Unusual Event because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In Cold Shutdown, the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RPV Inventory. Since the RCS usually remains intact in the Cold Shutdown Mode, a large inventory of water is available to keep the core covered. In Cold Shutdown, the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the Refueling Mode. Entry into Cold Shutdown conditions may be attained within hours of operating at power. Entry into the Refueling Mode procedurally may not occur for many hours after the reactor has been shut down. Thus, the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the Refueling Mode with irradiated fuel In the RPV (note that the heatup threat could be lower for Cold Shutdown conditions if entry into Cold Shutdown was following a refueling). In addition, the Operators should be able to monitor RCS temperature and RPV level so that escalation to the Alert level via CA2.1 or CA3.1 will occur if required. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 102 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] During refueling operations, the level in the RPV will normally be maintained above the vessel flange. Refueling evolutions that decrease water level below the vessel flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RCS temperatures depending on the time since shutdown. Normal means of core temperature indication and RPV level indication may not be available in the Refueling Mode. Redundant means of RPV level indication are therefore_ procedurally installed to assure the ability to monitor level will not be interrupted. Escalation to the Alert level via CA3.1 may therefore be required should an unplanned event result in RCS temperature exceeding the Technical Specification Cold Shutdown temperature limit for an extended period of time. The allowed time varies and is dependent on the status of the Primary Containment and Secondary Containment barriers and the integrity of the RCS barrier. The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. PROCEDURE 5.7.1 REVISION 67 PAGE 103 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV level indication for~ 15 min. (NOTE 3) NOTE 3- The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will _likely exceed the _applicable time.------------------------------------------------------------------------------- Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: RPV level is normally monitored using the following instruments (Reference 3, 4):

Wide Range NBI-Ll-85A, B & C (-155 to 60 inches).
Steam Nozzle Range NBI-Ll-92 (0 to 180 inches).
Fuel Zone Range NBI-Ll-91A, B & C (-320 to 60 inches).
Narrow Range RFC-Ll-94A, B & C (0 to 60 inches).
Shutdown Range NBI-Ll-86 (0 to 400 inches).

RPV level monitoring also includes the ability to monitor level visually in Refueling Mode consistent with escalation EAL CA2.1. Procedure 2.4RXLVL provides guidance for erratic or unexplained RPV water level changes. EOP/SAG Caution #1 indicates when an instrument may be used for level indication in the EOPs/SAGs. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 104 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Several Instruments are capable of providing Indication of RPV temperature with respect to the Technical Specification cold shutdown temperature limit (212°F). These include (Reference 1, 2):

NBI-TR-89, REACTOR VESSEL METAL TEMPERATURE RECORDER (Panel 9-21).
Vessel Drain, PMIS Point M180, or NBI-TR Point 06 if M180 is not available.
Vessel Bottom Head, PMIS Point M184, or NBI-TR Point 10 if M184 Is not available.
Bottom Head Adjacent to Support Skirt, PMIS Point M183, or NBI-TR Point 09.
RR-TR-165, RR SUCTION & FEEDWATER TEMP (Panel 9-4).

PMIS Points M174 through M185 can be used to monitor RPV temperatures. Thermocouples associated with computer Points M180, M183, and M185 do not respond as quickly nor register as high a temperature as other thermocouples due to their locations. CNS Basis Reference(s):

1. General Operating Procedure 2.1.1, Startup Procedure.

2. System Operating Procedure 2.2.69.2, RHR System Shutdown Operations.

3. Abnormal Procedure 2.4RXLVL, RPV Water Level Control Trouble.

4. Instrument Operating Procedure 4.6.1, Reactor Vessel Water Level Indication.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 105 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In Cold Shutdown, the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the RCS usually remains intact in the Cold Shutdown Mode, a large inventory of water is available to keep the core covered. In Cold Shutdown, the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the Refueling Mode. Entry into Cold Shutdown conditions may be attained within hours of operating at power. Entry into the Refueling Mode procedurally may not occur for many hours after the reactor has been shut down. Thus, the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the Refueling Mode with irradiated fuel in the RPV. Note that the heatup threat could be lower for Cold Shutdown conditions if entry into Cold Shutdown was following a refueling outage. In addition, the Operators should be able to monitor RCS temperature and RPV level so that escalation to the Alert level via CA2.1 or CA3.1 will occur if required. During refueling operations, the level in the RPV will normally be maintained above the vessel flange. Refueling operations that lower water level below the vessel flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RCS temperatures depending on the time since shutdown. Escalation to the Alert level via CA3.1 may therefore be required should an unplanned event result in RCS temperature exceeding the Technical Specification Cold Shutdown temperature limit for an extended period of time. The allowed time varies and is dependent on the status of the Primary Containment and Secondary Containment barriers and the integrity of the RCS barrier. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 106 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Unlike the Cold Shutdown Mode, normal means of core temperature indication and RCS level indication may not be available in the Refueling Mode. Redundant means of RPV level indication are therefore procedurally installed to assure the ability to monitor level will not be interrupted. However, if all level and temperature indication were to be lost in either the Cold Shutdown or Refueling Modes, this EAL would result in declaration of an Unusual Event if either temperature or level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be under CA2. l based on an inventory loss or CA3.1 based on exceeding its temperature criteria (212°F, Reference 3). The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. PROCEDURE 5.7.1 REVISION 67 PAGE 107 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert Any unplanned event results in EITHER: RCS temperature > 212°F for> Table C-3 duration (NOTE 4) OR RPV pressure increase > 10 psig due to a loss of RCS cooling NOTE 4 - Containment Closure is the action taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative _Procedure_ 0. 50. 5, O_utage _Shutdown_ Safety. _______________________________________________________________: Table.c*-~ .*RCS Reheat D1:,1ration.Thresholds.'

If an RCS heat removal system Is In operation within this time frame and RCS temperature is being reduced, the EAL is not applicable

1. RCS intact (Containment Closure NIA) 60 min.*

2. Containment Closure established AND 20 min.*

RCS not intact

3. Containment Closure not established AND 0min.

RCS not intact (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 108 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: A 10 psig RPV pressure increase can be read on: Primary:

Reactor pressure on PMIS Point 8025.
Reactor pressure on PMIS Point N013.
Reactor pressure on PMIS Point N014.

Backup: (Reference 3, 4)

RFC-PI-90A (Panel 9-5, 0 - 1200 psig).
RFC-PI-908 (Panel 9-5, 0 - 1200 psig).
RFC-PI-90C (Panel 9-5, 0 - 1200 psig).
NBI-PR-2A (Panel 9-3).
NBI-PR-28 (Panel 9-4).

Several instruments are capable of providing indication of RPV temperature with respect to the Technical Specification cold shutdown temperature limit (212°F). These include (Reference 1, 2):

NBI-TR-89, REACTOR VESSEL METAL TEMPERATURE RECORDER (Panel 9-21).
Vessel Drain, PMIS Point M180, or NBI-TR Point 06 if M180 is not available.
Vessel Bottom Head, PMIS Point M184, or NBI-,R Point 10 if M184 is not available.
Bottom Head Adjacent to Support Skirt, PMIS Point M183, or NBI-TR Point 09.
RR-TR-165, RR SUCTION & FEEDWATER TEMP (Panel 9-4).

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 109 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] PMIS Points M174 through M185 can be used to monitor RPV temperatures. Thermocouples associated with computer Points M180, M183, and M185 do not respond as quickly nor register as high a temperature as other thermocouples due to their locations. Inservice leak testing, hydrostatic testing, and control rod scram time testing in which RCS temperature is intentionally raised above 212°F per Technical Specification LCO 3.10.1 are not applicable to this EAL (Reference 5). CNS Basis Reference(s):

1. General Operating Procedure 2.1.1, Startup Procedure.

2. System Operating Procedure 2.2.69.2, RHR System Shutdown Operations.

3. Instrument Operating Procedure 4.6.2, Reactor Vessel Pressure Indication.

4. Emergency Procedure 5.9SAMG, Severe Accident Management Guidance, Technical Support Guidelines attachment (CPA TSG).

5. Technical Specifications LCO 3.10.1.

6. Technical Specifications Table 1.1-1.

NEI 99-01 Basis: The first condition of this EAL addresses events in which RCS temperature exceeds the CU3.1 EAL threshold of 212°F (Reference 6) for the durations identified in Table C-3. Table C-3, Duration #3, addresses complete loss of functions required for core cooling during Refueling and Cold Shutdown Modes when neither Containment Closure nor RCS integrity are established. RCS integrity is in place when the RCS pressure boundary is in its normal condition for the Cold Shutdown Mode of operation. No delay time is allowed for Duration #3 because the evaporated reactor coolant that may be released into the containment during this heatup condition could also be directly released to the environment. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 110 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Table C-3, Duration #2, addresses the complete loss of functions required for core cooling for > 20 minutes during Refueling and Cold Shutdown Modes when Containment Closure is established but RCS integrity is not established. RCS integrity should be assumed to be In place when the RCS pressure boundary is in its normal condition for the Cold Shutdown Mode of operation. The allowed 20 minute time frame was included to allow Operator action to restore the heat removal function, if possible. The allowed time frame is consistent with the guidance provided by Generic Letter 88-17, Loss of Decay Heat Removal, and is believed to be conservative given that a low pressure containment barrier to fission product release is established. The table note indicates that this duration is not applicable if actions are successful in restoring a RCS Heat Removal System to operation and RCS temperature is being reduced within the 20 minute time frame. Table C-3, Duration #1, addresses complete loss of functions required for core cooling for > 60 minutes during Refueling and Cold Shutdown Modes when RCS integrity is established. As in Durations #2 and #3, RCS integrity should be considered to be in place when the RCS pressure boundary is in its normal condition for the Cold Shutdown Mode of operation. The status of Containment Closure in this EAL is immaterial given the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety. The 10 psig pressure increase covers situations where, due to high decay heat loads, the time provided to restore temperature control should be < 60 minutes. The table note indicates that Duration #1 ls not applicable if actions are successful in restoring a RCS Heat Removal System to operation and RCS temperature is being reduced within the 60 minute time frame assuming the RCS pressure increase has remained < 10 psig. Escalation to Site Area would be via CS2.1 should boiling result in significant RPV level loss leading to core uncovery. A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary unplanned excursion above 212°F when the heat removal function is available and either the RCS is Intact or Containment Closure is established. The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. PROCEDURE 5. 7 .1 REVISION 67 PAGE 111 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 4 - Communications Initiating Condition: Loss of all on-site or off-site communications capabilities EAL: CU4.L Unusual Event Loss of all Table C-2 on-site (internal) communication methods affecting the ability to perform routine operations OR Loss of all Table C-2 off-site (external) communication methods affecting the ability to perform off-site notifications

                      --                                           ~

Table C-2 Communications* System*s - System Onsite Offsite (internal) (external) Station Intercom System "Gaitronics" X Site UHF Radio Consoles X Radio Paging System X Alternate Intercom X CNS On-Site Cell Phone System X X Telephone system (PBX) X X Federal Telecommunications System (FTS 2001) -x Local Telephones (C.O. Lines) X CNS State Notification Telephones X Satellite Telephones X Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, D - Defueled (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 112 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: NOTE - EPIP 5. 7COMMUN has more detail on each of the communications systems covered _by th is EAL.___________________________________________________________________________________________________ _ On-site/off-site communications include one or more of the systems listed in Table C-2 (Reference 1).

Station Intercom System "Gaitronics": Permits communication between the different parts of the plant and it also incorporates a public address system for plant wide announcements.
Site UHF Radio Consoles: The site UHF radio system uses four repeaters; Base 1 and Base 2 are used by Operations, Base 3 and Base 4 are used by Security. These repeaters operate on different frequencies. All remote control, portable, and mobile units are capable of selecting either repeater.
Radio Paging System: CNS leases pagers and radio paging services from a telecommunications company. Pagers are issued to various Management and Emergency Response personnel at CNS and other NPPD locations. Pagers can be activated from any touch-tone phone, on-site or off-site.
Alternate Intercom: Provides an alternate in-plant communications network utilizing a secondary telephone system. This system is located in the ERP shack and has battery back-up.
CNS On-Site Cell Phone System.
Telephone System (PBX): Provides voice communication between virtually all buildings, offices, and operation facilities within the station. The telephone system also provides communications between the plant and off-site facilities via the telephone switchboard network. The system allows Operating Crews to alert plant personnel in emergencies. The telephone company provides the normal and leased line services.
Federal Telecommunications System (FTS 2001): The Health Physics Network (HPN) and Emergency Notification System (ENS) provides communications between NRC and CNS during an emergency.
Local Telephones (C.O. Lines).

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 113 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

CNS State Notification Telephones: The CNS State Notification Telephone System is the primary means for the plant to make emergency notifications to state and local authorities. This system provides direct communication with the Nebraska State Patrol, the Missouri State Patrol, the Atchison County Sheriff's Department, and the Nemaha and Richardson County Sheriff's Departments.
Satellite Telephones.

This EAL is the cold condition equivalent of the hot condition EAL SU8.1. CNS Basis Reference(s):

1. EPIP 5.7COMMUN, Communications, Emergency Response Facility Communication Equipment attachment.

NEI 99-01 Basis: The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant Operations Staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with off-site authorities. The loss of off-site communications ability is expected to be significantly more comprehensive than the condition addressed by 10CFRS0. 72. The availability of one method of ordinary off-site communications is sufficient to inform state and local authorities of plant problems. This EAL is Intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being s~nt to off-site locations, etc.) are being utilized to make communications possible. The Table C-2 list for on-site communications loss encompasses the loss of all means of routine communications (e.g., commercial telephones, on-site cell phone systems, page party system, and radios/walkie talkies). The Table C-2 list for off-site communications loss encompasses the loss of all means of communications with federal off-site authorities. This should include the ENS, commercial telephone lines, telecopy transmissions, and dedicated phone systems. PROCEDURE 5.7.1 REVISION 67 PAGE 114 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 5 - Inadvertent Criticality Initiating Condition: Inadvertent criticality EAL: CU5.1 Unusual Event An unplanned sustained positive period observed on nuclear instrumentation Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: SRM A-D period Meters NMS-I-44A-D on Panel 9-5 identify this condition as well as Panel 9-5 amber light and SRM Period ( < 50 sec.) Annunciator 9-5-1/F-8 (Reference 1, 2). However, a SRM period alarm caused by SRM channel noise does not result in entry into this EAL (Reference 1). CNS Basis Reference(s):

1. Alarm Procedure 2.3_9-5-1, Panel 9 Annunciator 9-5-1, F-8, SRM Period.

2. Instrument Operating Procedure 4.1.1, Source Range Monitoring System.

NEI 99-01 Basis: This EAL addresses criticality events that occur in Cold Shutdown or Refueling Modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States) such as fuel mis-loading events. This EAL indicates a potential degradation of the level of safety of the plant, warranting an Unusual Event classification. This EAL excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated) which are addressed in the companion EAL SU2.1. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 115 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] The term "sustained" is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alteration. These short term positive periods are the result of the increase in neutron population due to subcritical multiplication. Escalation to higher classification levels would be by the judgment EALs In Category H (EAL HA6.1, HS6.1, or HG6.1). PROCEDURE 5. 7 .1 REVISION 67 PAGE 116 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 6 - Loss of DC Power Initiating Condition: Loss of required DC power for 15 minutes or longer EAL: CU6. l Unusual Event

 < 105 VDC bus voltage indications on all Technical Specification required 125 VDC buses for~ 15 min. (NOTE 3)

NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it ls determined the condition will Ii kely exceed the_a p pl ica ble time. ______________________________________________________________________________ _ Mode Applicability: 4 - Cold Shutdown, 5 - Refueling CNS Basis: 105 VDC is the minimum design bus voltage (Reference 4). The 125 VDC System supplies DC power to conventional station emergency equipment and selected Safeguard System loads. 125 VDC Distribution Panels supply control and instrument power for annunciators control logic power and protective relaying. Figure C-2 illustrates the 125 VDC Power System (Reference 3). If 125 VDC Distribution Panel A is lost, the following major equipment is affected: RRMG A speed and breaker control, 4160V Bus lA, lE, and lF breaker control and undervoltage logics, 480V Bus lA and lF breaker control, the right light in all Control Room annunciators, annunciator panels for Water Treatment, RHR A Gland Water, Auxiliary Steam Boiler C, DG-1 starting and breaker control logics, CS A, RCIC, and RHR A control logics, TIP valve control monitors, main generator voltage regulation, RFPT A trip logic, and ARI solenoid valve power. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 117 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If 125 VDC Distribution Panel B is lost, the following major equipment is affected: RRMG B speed and breaker control, 4160V Bus lB and lG breaker control and undervoltage logics, 480V Bus lB and lG breaker control, the left light in all Control Room annunciators, annunciator panels for ALRW, RHR B Gland Water, Auxiliary Steam Boiler D, DG-2 starting and breaker control logics, CS B, HPCI, and RHR B control logics, main generator trip logic, main generator and transformer protective relaying, bypass valves fail to control pressure after turbine trip, and RFPT B trip logic. Battery chargers receive their power from 460V critical motor control centers. Each 125 VDC bus receives power from either a 125 VDC battery or a 125 VDC battery charger. The battery chargers receive their power from 460V critical motor control centers. The 250 VDC System supplies DC power to conventional station emergency equipment and selected Safeguard System loads. Although 250 VDC Buses lA and lB provide vital DC emergency power, 250 VDC Safety System loads (such as motor operated valves) also require 125 VDC control power. Loss of 125 VDC buses alone, therefore, would render most Safeguard System loads inoperable (Reference 4, 5, 6). This EAL is the cold condition equivalent of the hot condition loss of DC power EAL 557.1. CNS Basis Reference(s):

1. Emergency Procedure 5.3DC125, Loss of 125 VDC.

2. Surveillance Procedure 6.EE.607, 125V Station Battery Modified Performance Discharge Test.

3. BR 3058 DC One Line Diagram.

4. Technical Specifications B 3.8.4.

5. USAR Section Vlll-6.2.

6. USAR Section VIII-6.3.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 118 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure C-2: 125 VDC Power System MCC-L MCC-TX

                                           ---~-

TX I 12SVOC I .---------. I I .....,......,...,..___,....., I I I I I I I I GND I MCCI-TX

Ll~~s:

I L __:: _::~ _::]-- - _:c _- - -- - - I I

                                                                                            ~'~l---

I I I

                                      ------t                         I           :

I I I I BATTERY BATTERY I ET I 1A 1B I GND

                                                                         -            D&T

GHTS: ":' -:- I LIG I

L_ A I I A I I I I I I I I I I I I L.------ N E RCIC E c" N RxBLDG N OE DIST. PANEL A

                                                                                                ,121SV DC SR "B*

e N 6 DIST. PANEL B HPCI tt HPCI PANEL M0-16 continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 119 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the Operating Crew may not have necessary indication and control of equipment needed to respond to the loss. The plant will routinely perform maintenance on a train related basis during shutdown periods. The required buses are the minimum allowed by Technical Specifications for the mode of operation. It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be via CA3. l. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. PROCEDURE 5.7.1 REVISION 67 PAGE 120 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL: HU 1.1 Unusual Event Seismic event identified by any two of the following:

The Seismic Monitor System free field sensor actuated or Alarm B-3/B-1, SEISMIC EVENT
Earthquake felt in plant
National Earthquake Information Center Mode Applicability:

All CNS Basis: The method of detection with respect to emergency classification relies on the agreement of the Shift Operators on-duty in the Control Room that the suspected ground motion is a "felt earthquake" as well as the actuation of the CNS seismic instrumentation. Consensµs of the Control Room Operators with respect to ground motion helps avoid unnecessary classification if the seismic switches inadvertently trip or detect vibrations not related to an earthquake. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 121 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS seismic instrumentation actuates at 0.01 g. The free field seismic sensor, located in the metal enclosure north of the Intake Structure, provides a start signal to the Seismic Monitoring System when ground motion > 0.01 g is sensed. On receipt of this start signal, the Seismic Monitor System indicates or initiates:

An Event is in-progress.
An Event has been recorded.
Annunciator B-3/B-1, SEISMIC EVENT, alarms.
Recording of the seismic signals from sensors located north of the Intake Structure, in the Reactor Building NW Quad 859', and on the Reactor Building 1001' level.

The National Earthquake Information Center can confirm if an earthquake has occurred in the area of the plant. Refer to the National Earthquake Information Center website: http://earthquake.usgs.gov/eqcenter/ This event escalates to an Alert under EAL HAL 1 if the earthquake exceeds Operating Basis Earthquake (QBE) levels. In the absence of operable seismic switches, the Earthquake Magnitude Correlation below may be used to perform an assessment of the relative magnitude of an earthquake. ( continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 122 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Modified Merica Iii Maximum Richter Intensity Acceleration Magnitude Scale Description of Effects* (g) (Descriptor) Not felt except by a very few under especially favorable I. 1.0 to 2.9 conditions. 0.001 to (Very Felt only by a few persons at rest, especially on upper floors 0.002 II. Minor) of buildings. Felt quite noticeably by persons Indoors, especially on upper floors of buildings. Many people do not recognize It as an III. earthquake. Standing motor cars may rock slightly. 0.003 to 3.0 to 3.9 Vibrations similar to the passing of a truck. Duration 0.010 (Minor) estimated. Felt indoors by many, outdoors by few during the day. At night, some awakened. Dishes, windows, doors disturbed; IV. walls make cracking sound. Sensation like heavy truck striking building. Standing motor cars rocked noticeably. 0.011 to 4.0 to 4.9 Felt by nearly everyone; many awakened. Some dishes, 0.029 (Light) V. windows broken. Unstable objects overturned. Pendulum clocks may stop. Felt by all, many frightened. Some heavy furniture moved; 0.03 to 0.2 VI. a few Instances of fallen plaster. Damage slight. (CNS Design Damage negligible in buildings of good design and Basis 5.0 to 5.9 construction; slight to moderate in well-bullt ordinary OBE = 0.1; (Moderate) VII. structures; considerable damage In poorly built or badly CNS Design designed structures; some chimneys broken. Basis SSE= 0.2) Damage slight in specially designed structures; considerable damage in ordinary substantial buildings with partial VIII. collapse. Damage great in poorly built structures. Fall of 6.0 to 6.9 chimneys, factory stacks, columns, monuments, walls. 0.2 to 0.4 (Strong) Heavy furniture overturned. Damage considerable In specially designed structures; well-designed frame structures thrown out of plumb. IX. Damage great In substantial buildings, with partial collapse. Buildings shifted off foundations. 7.0 to 7.9 0.4 to 0.5 Some well-built wooden structures destroyed; most masonry (Major)

x. and frame structures destroyed with foundations. Ralls bent.

Few, if any (masonry), structures remain standing. Bridges XI. destroyed. Rails bent greatly. 8.0 and

                                                                              > 0.5          higher Damage total. Lines of sight and level are distorted.                           (Great)

XII. Objects thrown Into the air.

The Richter magnitude and maximum acceleration (g) In the table above is provided for general comparison and description. Actual seismic intensity and effects will vary.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 123 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Alarm Procedure 2.3_8-3, Panel B - Annunciator B-3/B-1.

2. Alarm Procedure 2.3_8-3, Panel B - Annunciator B-3/A-1.

3. Instrument Operating Procedure 4.12, Seismic Instrumentation.

4. Emergency Procedure 5. lQuake, Earthquake.

5. USAR Section Il-5.2.4 and Table Il-5-1.

NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant Operators. Damage may be caused to some portions of the site, but should not affect ability of safety functions to operate. As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion Is felt at the nuclear plant site and recognized as an earthquake based on a consensus of Control Room Operators on-duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. The National Earthquake Information Center can confirm if an earthquake has occurred in the area of the plant. PROCEDURE 5.7.1 REVISION 67 PAGE 124 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL: HUl.2 Unusual Event Tornado striking within Protected Area boundary OR Sustained high winds~ 100 mph Mode Applicability: All CNS Basis: 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. The design wind pressure for the station and structures is 30 lb/ft2 which is equivalent of sustained winds up to 100 mph (Reference 3). The Protected Area refers to the designated security area around the process buildings. Sustained winds are of a prolonged duration and, therefore, do not include gusts. Sustained winds are not intermittent or of a transitory nature. Since the inauguration of the Automatic Surface Observation System (ASOS), the National Weather Service has adopted a 2 minute average standard for its sustained wind definition (Reference 1). (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 125 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. National Weather Service webpage "http://www.aoml.noaa.gov/hrd/tcfaq/D4.h tml".

2. USAR Section II-3.2.2.

NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant Operators. This EAL is based on a tornado striking (touching down) or high winds within the Protected Area. Escalation of this emergency classification level, if appropriate, would be based on visible damage or by other in plant conditions via HAl.2. PROCEDURE 5.7.1 REVISION 67 PAGE 126 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL: HUl.3 Unusual Event Main turbine failure resulting in casing penetration or damage to turbine or generator seals Mode Applicability: All CNS Basis: The main turbine generator stores large amounts of rotational kinetic energy in its rotor. In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those housing safety-related equipment. In the event of missile ejection, the probability of a strike on a plant region is a function of the energy and direction of an ejected missile and of the orientation of the turbine with respect to the plant region. In addition to potential missile generation, the failure of the rotational elements will cause imbalances in these elements that may be of sufficient magnitude to damage turbine and generator seals resulting in the release of large quantities of lube oil and/or hydrogen. CNS Basis Reference(s): None (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 127 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant Operators. This EAL addresses main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual fires and flammable gas build up are appropriately classified via HU2. l and HU3.1. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety-related equipment. Escalation of this emergency classification level, if appropriate, would be to HAl.3 based on damage done by projectiles generated by the failure or by any radiological releases. These latter events would be classified by the Category A, Abnormal Rad Release/Rad Effluent EALs, or the Category F, Fission Product Barrier EALs. PROCEDURE 5.7.1 REVISION 67 PAGE 128 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL: HU1.4 Unusual Event Flooding In any Table H-1 area that has the potential to affect safety-related equipment required by Technical Specifications for the current operating mode

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

SLC _(Stand by_ Liquid Control), SW_ (Service_ Water).-------------------------------------------------- Mode Applicability: All (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 129 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: The internal flooding areas of concern are listed in Table H-1 (Reference 1-5). Flooding, as used in this EAL, describes a condition where water is entering the room faster than installed equipment is capable of removal, resulting in a rise of water level within the room. Flooding in these areas could have the potential to cause a reactor trip and could result In consequential failures to important systems. The potential for flooding in these areas was determined by an examination of piping systems in the area and also considered propagation of water from one area to another. The accumulation of water resulting in a rising water level in the area constitutes flooding. The Critical Switchgear Rooms are a part of the Reactor Building. CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-FP-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section XIl-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant Operators. This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. Escalation of this emergency classification level, if appropriate, would be based visible damage via HAl.4 or by other plant conditions. PROCEDURE 5.7.1 REVISION 67 PAGE 130 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL: HUl.5 Unusual Event High river/forebay water level > 899' MSL OR Low river level/forebay < 870' MSL Mode Applicability: All CNS Basis: This EAL covers high river/forebay water level conditions that could be a precursor of more serious events as well as low river/forebay water level conditions which may threaten operability of plant cooling systems. 899' MSL is the level associated with the maximum flood of record (Reference 1). 870' MSL is the Minimum Probable river level (Reference 2). A further level drop may threaten availability of cooling systems and heat sink. The forebay refers to the area between the east Intake Structure wall and the guide wall (Reference 3). CNS Basis Reference(s):

1. USAR Section II-4.2.2.2.

2. USAR Section II-4.2.3.1.

3. USAR Section II-4.2.3.2.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 131 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant Operators. This EAL addresses other site specific phenomena (such as flood) that can also be precursors of more serious events. PROCEDURE 5.7.1 REVISION 67 PAGE 132 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL: HAl.1 Alert Seismic event > 0.1 g as indicated by the Seismic Monitor System free field sensor or Alarm B-3/A-1, EMERGENCY SEISMIC HIGH LEVEL AND Earthquake confirmed by any of the following:

Earthquake felt in plant
National Earthquake Information Center
Control Room indication of degraded performance of systems required for the safe shutdown of the plant Mode Applicability:

All CNS Basis: CNS seismic instrumentation actuates at 0.01 g. The free field seismic sensor, located in the metal enclosure north of the Intake Structure, provides a start signal to the Seismic Monitoring System when ground motion > 0.01 g is sensed. On receipt of this start signal, the Seismic Monitor System indicates or initiates:

An Event is in-progress.
An Event has been recorded.
Annunciator B-3/B-1, SEISMIC EVENT, alarms.
Recording of the seismic signals from sensors north of the Intake Structure, in the Reactor Building NW Quad 859', and on the Reactor Building 1001' level are recorded.
Alarm B-3/A-1, EMERGENCY SEISMIC HIGH LEVEL, is received if the seismic activity exceeds O.1 g.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 133 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] The National Earthquake Information Center can confirm if an earthquake has occurred in the area of the plant. Refer to the National Earthquake Information Center website: http://earthquake.usgs.gov/eqcenter/ In the absence of operable seismic switches, the Earthquake Magnitude Correlation below may be used to perform an assessment of the relative magnitude of an earthquake. An earthquake in excess of the CNS Operating Basis Earthquake (QBE) levels should be classified under this EAL. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 134 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Modified Merica Iii Maximum Richter Intensity Acceleration Magnitude Scale Description of Effects* (g) (Descriptor) Not felt except by a very few under especlally favorable I. 1.0 to 2.9 conditions. 0.001 to (Very Felt only by a few persons at rest, especlally on upper floors 0.002 II. Minor) of buildings. Felt quite noticeably by persons Indoors, especlally on upper floors of buildings. Many people do not recognize it as an III. earthquake. Standing motor cars may rock sllghtly. 0.003 to 3.0 to 3.9 Vibrations similar to the passing of a truck. Duration 0.010 (Minor) estimated. Felt indoors by many, outdoors by few during the day. At night, some awakened. Dishes, windows, doors disturbed; N. walls make cracking sound. Sensation like heavy truck striking building. Standing motor cars rocked noticeably. 0.011 to 4.0 to 4.9 Felt by nearly everyone; many awakened. Some dishes, 0.029 (Light) V. windows broken. Unstable objects overturned. Pendulum clocks may stop. Felt by all, many frightened. Some heavy furniture moved; 0.03 to 0.2 VI. a few Instances of fallen plaster. Damage slight. (CNS Design Damage negllglble In buildings of good design and Basis 5.0 to 5.9 construction; slight to moderate in well-built ordinary OBE = 0.1; (Moderate) VII. structures; considerable damage In poorly bullt or badly CNS Design designed structures; some chimneys broken. Basis SSE= 0.2) Damage slight In specially designed structures; considerable damage in ordinary substantial bulldings with partial VIII. collapse. Damage great in poorly bullt structures. Fall of 6.0 to 6.9 chimneys, factory stacks, columns, monuments, walls. 0.2 to 0.4 (Strong) Heavy furniture overturned. Damage considerable in specially designed structures; well-designed frame structures thrown out of plumb. IX. Damage great in substantial buildings, with partial collapse. Buildings shifted off foundations. 7.0 to 7.9 0.4 to 0.5 Some well-built wooden structures destroyed; most masonry (Major) X. and frame structures destroyed with foundations. Ralls bent. Few, If any (masonry), structures remain standing. Bridges XI. destroyed. Ralls bent greatly. 8.0 and

                                                                              > 0.5         higher Damage total. Lines of sight and level are distorted.                          (Great)

XII. Objects thrown into the air.

The Richter magnitude and maximum acceleration (g) in the table above is provided for general comparison and description. Actual seismic intensity and effects will vary.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 135 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Alarm Procedure 2.3_B-3, Panel B - Annunciator B-3/A-1.

1. Alarm Procedure 2.3_B-3, Panel B - Annunciator B-3/B-1.

2. Instrument Operating Procedure 4.12, Seismic Instrumentation.

3. Emergency Procedure 5.lQuake, Earthquake.

4. USAR Section 11-5.2.4 and Table 11-5-1.

NEI 99-01 Basis: This EAL escalates from HU1.1 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other EALs (e.g., System Malfunction). Seismic events of this magnitude can result in a Vital Area being subjected to forces beyond design limits, and thus, damage may be assumed to have occurred to plant safety systems. The National Earthquake Information Center can confirm if an earthquake has occurred in the area of the plant. PROCEDURE 5. 7 .1 REVISION 67 PAGE 136 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL: HAl.2 Alert Tornado striking or high winds ~ 100 mph resulting in EITHER: Visible damage to any Table H-1 area structure containing safety systems or components OR Control Room indication of degraded performance of safety systems Table H-1 S~fe Shutdown Areas,

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual

Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

! SLC _(Standby_ Liquid_ Control), SW _(Service _Water).--------------------------------------------------- Mode Applicability: All (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 137 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: This threshold addresses events that may have resulted in Safe Shutdown Areas being subjected to forces (tornado or high winds > 100 mph) (Reference 6) beyond design limits. Table H-1 safe shutdown areas house equipment the operation of which may be needed to ensure the reactor safely reaches and is maintained shutdown (Reference 1-4, 7). A tornado striking (touching down) within the Protected Area resulting in visible damage warrants declaration of an Alert 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. The design wind pressure for the station and structures is 30 lb/ft2 which is the equivalent of sustained winds up to 100 mph (Reference 6). The Protected Area refers to the designated security area around the process buildings. The Critical Switchgear Rooms are a part of the Reactor Building. CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-FP-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section II-3.2.2.

6. USAR Section XII-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 138 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL escalates from HU1.2 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other EALs (e.g., System Malfunction). This EAL is based on a tornado striking (touching down) or high winds that have caused visible damage to structures containing functions or systems required for safe shutdown of the plant. PROCEDURE 5. 7 .1 REVISION 67 PAGE 139 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL: HA1.3 Alert Main turbine failure-generated projectiles result in EITHER: Visible damage to or penetration of any Table H-1 area structure containing safety systems or components OR Control Room indication of degraded performance of safety systems Table H-1 Safe Shutdown Areas

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

_SLC _(Standby_ Liquid .Control), SW _(Service_Water). -------------------------------------------------- Mode Applicability: All (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 140 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: The main turbine generator stores large amounts of rotational kinetic energy in its rotor. In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those housing safety-related equipment. In the event of missile ejection, the probability of a strike on a plant region is a function of the energy and direction of an ejected missile and of the orientation of the turbine with respect to the plant region. The list of Table H-1 areas includes all areas containing safety-related equipment, their controls, and their power supplies (Reference 1-5). The Critical Switchgear Rooms are a part of the Reactor Building. CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-F P-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section XII-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 141 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The EAL escalates from HU1.3 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other EALs (e.g., System Malfunction). PROCEDURE 5.7.1 REVISION 67 PAGE 142 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL: HAl.4 Alert Flooding in any Table H-1 area resulting in EITHER: An electrical shock hazard that precludes access to operate or monitor safety equipment OR Control Room indication of degraded performance of safety systems Table ,H:1 Safe Shutdown Areas

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

SLC _(Stand by_ Liquid Control), SW_ (Service. Water).-------------------------------------------------- Mode Applicability: All (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 143 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: The internal flooding areas of concern are listed in Table H-1 (Reference 1-5). Flooding In these areas could have the potential to cause a reactor trip and could result in consequential failures to important systems.. The potential for flooding in this area was determined by an examination of piping systems in the area and also considered propagation of water from one area to another. The accumulation of water resulting in a rising water level in the area constitutes flooding. The Critical Switchgear Rooms are a part of the Reactor Building. CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-FP-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section XII-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 144 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The EAL escalates from HUl.4 in that the occurrence of the event has resulted in an electrical shock hazard precluding access to plant structures or areas containing equipment necessary for a safe shutdown or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not, be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other EALs (e.g., System Malfunction). This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. It is based on the degraded performance of systems or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to access, operate, or monitor safety equipment represents an actual or substantial potential degradation of the level of safety of the plant. Flooding as used in this EAL describes a condition where water is entering the room faster than installed equipment is capable of removal, resulting in a rise of water level within the room. Classification of this EAL should not be delayed while corrective actions are being taken to isolate the water source. PROCEDURE 5. 7.1 REVISION 67 PAGE 145 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL: HAl.5 Alert High river/forebay water level > 902' MSL OR Low river/forebay level < 865' MSL Mode Applicability: All CNS Basis: HUl.5 covers high river/forebay water level conditions that could pose a significant threat to plant safety as well as low river/forebay water level conditions which may threaten operability of vital emergency plant cooling systems. A river level of 901' 8" requires reactor shutdown and a river level of 902' represents the maximum possible (10,000 year) flood stage (Reference 1, 3). A river level of 865' MSL corresponds to the Safe Shutdown low river level and threatens availability of cooling systems and heat sink (Reference 4). The forebay refers to the area between the east Intake Structure wall and the guide wall (Reference 2). CNS Basis Reference(s):

1. Emergency Procedure 5. lFLOOD, Flood.

2. USAR Section II-4.2.2.1.

3. USAR Section II-4.2.2.2.

4. USAR Section II-4.2.3.2.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 146 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Escalation to higher classifications occurs on the basis of other EALs (e.g., System Malfunction). This EAL addresses other site specific phenomena that result in visible damage to Vital Areas or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant (such as flood) that can also be precursors of more serious events. PROCEDURE 5. 7 .1 REVISION 67 PAGE 147 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL: HA1.6 Alert Vehicle crash resulting in EITHER: Visible damage to any Table H-1 area structure containing safety systems or components OR Control Room indication of degraded performance of safety systems Table H-1 Safe Shutdowr.a Areas

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

_SLC _(Stand by_ Liquid Control), SW_ (Service_ Water).-------------------------------------------------- Mode Applicability: All (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 148 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: Table H-1 Safe Shutdown Areas house equipment the operation of which may be needed to ensure the reactor reaches and is maintained in shutdown (Reference 1-4, 6). The Protected Area refers to the designated security area around the process buildings. If vehicle crash ls determined to be hostile in nature, the event is classified u'nder security based EALs. The Critical Switchgear Rooms are a part of the Reactor Building. CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-FP-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section XIl-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 149 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is based on the occurrence of a vehicle crash that has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown or has caused damage to the safety systems in those structures evidenced by control Indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response Is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other EALs (e.g., System Malfunction). This EAL addresses vehicle crashes within the PROTECTED AREA that resul,ts In VISIBLE DAMAGE to VITAL AREAS or indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant. PROCEDURE 5. 7 .1 REVISION 67 PAGE 150 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 2 - Fire or Explosion Initiating Condition: Fire within the Protected Area not extinguished within 15 minutes of detection or explosion within the Protected Area EAL: HU2.1 Unusual Event Fire in any Table H-1 area not extinguished within 15 min. of Control Room notification or receipt of a valid Control Room alarm due to fire (NOTE 3) NOTE 3- The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely exceed the_a pp Iica bl e ti me. __ ---------------------------------- __________________________________________ _ Table H-1 Safe Shutdown Areas

Reactor Building
Conttol Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room Mode Applicability:

All CNS Basis: Fire, as used in this EAL, means 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. The Critical Switchgear Rooms are a part of the Reactor Building. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 151 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVE:L TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-FP-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section XII-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

NEI 99-01 Basis: This EAL addresses the magnitude and extent of fires that may be potentially significant precursors of damage to safety systems. It addresses the fire and not the degradation in performance of affected systems that may result. As used here, detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a fire is occurring or indication of a fire detection system alarm/actuation. Verification of a fire detection system alarm/actuation includes actions that can be taken within the Control Room or other nearby site specific location to ensure it is not spurious. An alarm is assumed to be an indication of a fire unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm. The intent of this 15 minute duration is to size the fire and to discriminate against small fires that are readily extinguished (e.g., smoldering waste paper basket). PROCEDURE 5. 7 .1 REVISION 67 PAGE 152 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 2 - Fire or Exploslon Initiating Condition: Fire within the Protected Area not extinguished within 15 minutes of detection or explosion within the Protected Area EAL: HU2.2 Unusual Event Explosion within the Protected Area Mode Applicability: All CNS Basis: The Protected Area refers to the designated Security Area around the process buildings. As used here, an explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials. A steam line break or steam explosion that damages surrounding permanent structures or equipment would be classified under this EAL. This does not mean the emergency is classified simply because the steam line break occurred. The method of damage is not as important as the degradation of plant structures or equipment. The need to classify the steam line break itself is considered in fission product barrier degradation monitoring (EAL Category F). If explosion is determined to be hostile in nature, the event is classified under security based EALs. CNS Basis Reference(s): None (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 153 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL addresses the magnitude and extent of explosions that may be potentially significant precursors of damage to safety systems. It addresses the explosion and not the degradation in performance of affected systems that may result. As used here, detection is visual observation and report by plant personnel or sensor alarm indication. This EAL addresses only those explosions of sufficient force capable of causing damage to permanent structures or equipment within the Protected Area. No attempt is made to assess the actual magnitude of any damage. The occurrence of the explosion is sufficient for declaration. The Emergency director also needs to consider any security aspects of the explosion, if applicable. Escalation of this emergency classification level, if appropriate, would be based on HA2.l. PROCEDURE 5. 7 .1 REVISION 67 PAGE 154 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 2 - Fire or Explosion Initiating Condition: Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown EAL: HA2.1 Alert Fire or explosion resulting in EITHER: Visible damage to any Table H-1 area containing safety systems or components OR C,ontrol Room indication of degraaed performance of safety systems l"a.ble H4 .Safe -Shutdov/n Areas

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary

! Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual ! Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

SLC (Standby Liquid Control), SW (Service Water).

~------------------------------------------------------------------------------------------------------------------------------------ Mode Applicability: All (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 155 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: Fire, as used in this EAL, means 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. An explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials. A steam line break or steam explosion that damages permanent structures or equipment would be classified under this EAL. The method of damage is not as important as the degradation of plant structures or equipment. The need to classify the steam line break itself is considered in fission product barrier degradation monitoring (EAL Category F). The Critical Switchgear Rooms are a part of the Reactor Building. CNS Basis Reference(s):

1. Site Services Procedure 1.1, Station Security.

2. CNS-FP-60, Fire Area Boundary Drawing Index.

3. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

4. Fire Safety Analysis Calculations.

5. USAR Section XII-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 156 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Visible damage is used to identify the magnitude of the fire or explosion and to discriminate against minor fires and explosions. The reference to structures containing safety systems or components is included to discriminate against fires or explosions in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact the fire or explosion was large enough to cause damage to these systems. The use of visible damage should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform detailed damage assessments. The Emergency Director also needs to consider any security aspects of the explosion. Escalation of this emergency classification level, if appropriate, will be based on System Malfunctions, Fission Product Barrier Degradation, or Abnormal Rad Levels/Radiological Effluent EALs. PROCEDURE 5. 7.1 REVISION 67 PAGE 157 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 3 - Hazardous Gas

Initiating Condition: Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to normal plant operations EAL: HU3.1 Unusual Event Toxic, corrosive, asphyxiant, or flammable gases in amounts that have or could affect normal plant operations Mode Applicability: All CNS Basis: As used in this EAL, affecting normal plant operations means that activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures have been impacted. Entry into Abnormal or Emergency Operating Procedures, or deviation from normal security or radiological controls posture is a dep 9rture from normal plant operations, and thus, would be considered to have been affected. Administrative Procedure 0.36.6, Monitoring for Industrial Gases, may be used for help in assessing this EAL. Such review, however, does not constitute a departure from normal operations. The release may have originated within the Site Boundary or it may have originated off-site and subsequently drifted onto the Site Boundary. Off-site events (e.g., tanker truck accident releasing toxic gases, etc.) resulting in the plant being within the evacuation area should also be considered in this EAL because of the adverse effect on normal plant operations. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 158 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] At CNS there are various potential sources of atmospheric contamination. Some of these sources are;

Inert gas used for oxygen exclusion (nitrogen).
Combustion products.
Carbon dioxide from fire extinguishing.
Halon from fire extinguishing.
Welding gases (enclosed areas).
Vapors from painting (enclosed areas).
Vapors from petroleum products.
Hydrogen (OWC Hydrogen Gas Generation System, Generator Cooling System, batteries, and disassociation of water in the reactor).
Asphyxiants and irritants, found most often in confined areas (water and oil storage tanks, open manholes).
Methane from bacterial action (tanks and pits).

Some of the gases which could affect normal plant operations under this EAL are:

Carbon Monoxide - One of the most common asphyxiants encountered in industry. It is formed by the incomplete combustion of fuel containing carbon.

It may be found in the vicinity of a fire or a leak in an exhaust system (flue gas or internal combustion engines).

Oxygen - Oxygen has two fundamentally important properties: it supports combustion and it supports life. Since oxygen is necessary for life, it must be present in sufficient quantity. Oxygen deficiency occurs in confined spaces where the level of oxygen has been reduced below the limit to support life.

Oxygen content in the air can become fatally low in a brief period of time. Some of the more common causes of this problem are oxidation of metals, bacterial action, combustion, and displacement by other gases. An enriched oxygen atmosphere will accelerate combustion. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 159 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

Hydrogen - Used in generator cooling. Hydrogen gas is also produced by the owe Gas Generation System, located in the owe Building, and subsequently injected Into the Condensate System just upstream of the condensate booster pumps. Hydrogen is also produced by the disassociation of water from radiation in the reactor, which is seen in the off-gas. The presence of hydrogen will be especially significant in the Off-Gas and Augmented Off-Gas Systems.

Hydrogen is also a by-product of battery charging. It is lighter than air so it will be found in pockets at the ceiling of enclosures.

Argon - Commonly used during the welding of certain metals. It is denser than air so it will settle in pockets below the welding area.
Carbon Dioxide - Used to fight fire. Being heavier than air, carbon dioxide will settle in pockets and displace oxygen.
Halon - Used to fight fire in the Service Water Pump Room, Computer Room, and in the Simulator. Discharge of a Halon System will result in exceeding IDLH limits in the area. Discharge of the Service Water Pump Room or Computer Room Halon System should be classified under EAL HA3.1.
Nitrogen - Used primarily to purge Primary Containment. Since it is approximately the same density as air, it can be dispersed by proper ventilation. Areas of poor ventilation may contain greater than expected concentrations of nitrogen and consequently may be deficient in oxygen.
Combustible Gases and Vapors - Includes naturally occurring gases (such as methane and hydrogen gas) and the vapors of a large group of liquids which are used as fuels and solvents. Monitoring shall be required in fuel tanks and other areas where explosive mixtures may be present.
Hydrogen Sulfide - Classified as an irritant in low concentrations, but is even more toxic than carbon monoxide because It inflames the mucus membranes and results in the lungs filling with fluid. This colorless gas has a characteristic rotten egg odor, which renders the sense of smell ineffective. Hydrogen sulfide may be found in sewage treatment or wherever organic matter containing sulfur decomposes and shall be monitored constantly during work.
Methane - The chief constituent of natural gas and is extremely explosive. It ls non-toxic, but may reduce the oxygen content of an atmosphere, causing asphyxiation. Methane is often found in the vicinity of sanitary landfills and has been detected In tanks where bacterial action is taking place (i.e., reactor water cleanup and condensate phase separator tanks). It is lighter than air and tends to accumulate in high spots or pockets. This can present a dangerous situation in storage tanks or sewers where access ls normally gained at the top of the confined area.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 160 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

Ethyl Benzene - Used primarily as an additive to diesel fuel. Acute exposure results in a local irritant effect on the skin and mucous membranes. Chronic exposure can lead to nervous system disorders and upper respiratory tract Inflammation. Monitoring is required when entering a diesel fuel tank.
Chlorine - Used in chemical treatment of Circulating Water and Service Water Systems. Chlorine gas can be recognized by its pungent, irritating odor, which is like the odor of bleach. Chlorine Is not flammable but can react explosively with other chemicals such as turpentine or ammonia. Chlorine gas stays close to the ground and spreads rapidly. When chlorine gas comes in contact with moist human tissues, such as the eyes throat and lungs, an acid is produced that can damage these tissues.
Chlorine Dioxide -'This is a yellow to reddish-yellow manufactured gas which does not occur naturally in the environment. When added to water, chlorine dioxide forms chlorite ion, which is also a very reactive chemical. High levels of chlorine dioxide can be Irritating to the nose, eyes, throat, and lungs.
Hydrogen Chloride - This is a colorless to slightly yellowish gas with a pungent odor. On exposure to air, the gas forms dense white vapors due to condensation with atmospheric moisture. The vapor is corrosive and air concentrations above 5 ppm can cause irritation. When mixed with water or atmospheric moisture, a highly corrosive atmosphere is formed. The most common source of Hydrogen Chloride gas is from Muriatic (Hydrochloric) Acid.

Should the release affect access to plant Safe Shutdown Areas, escalation to an Alert would be based on EAL HA3.1. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1. CNS Basis Reference(s):

1. Administrative Procedure 0.36.6, Monitoring for Industrial Gases.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 161 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is based on the release of toxic, corrosive, asphyxiant, or flammable gases of sufficient quantity to affect normal plant operations. The fact that SCBA may be worn does not eliminate the need to declare the event. This EAL is not intended to require significant assessment or quantification. It assumes an uncontrolled process that has the potential to affect plant operations. This would preclude small or incidental releases or releases that do not impact structures needed for plant operation. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness, or even death. Escalation of this emergency classification level, if appropriate, would be based on HA3.1. PROCEDURE 5. 7.1 REVISION 67 PAGE 162 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 3 - Toxic, Corrosive, Asphyxiant, And Flammable Gas Initiating Condition: Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to normal plant operations EAL: HU3.2 Unusual Event Recommendation by local, county, or state officials to evacuate or shelter site personnel based on an off-site event Mode Applicability: All CNS Basis: This EAL is based on the existence of an uncontrolled release originating off-site and local, county, or state officials have reported the need for evacuation or sheltering of site personnel. Off-site events (e.g., tanker truck accident releasing toxic gases, etc.) are considered in this EAL because they may adversely affect normal plant operations. State officials may determine the evacuation area for off-site spills by using the Department of Transportation (DOT) Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. If evacuation area extends to any portion of the Owner Controlled Area, the EAL threshold is met. Should the release affect plant Safe Shutdown Areas, escalation to an Alert would be based on EAL HA3.1. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1. CNS Basis Reference(s):

1. Administrative Procedure 0.36.6, Monitoring for Industrial Gases.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 163 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is based on the release of toxic, corrosive, asphyxlant, or flammable gases of sufficient quantity to affect normal plant operations. The fact that SCBA may be worn does not eliminate the need to declare the event. This EAL is not intended to require significant assessment or quantification. It assumes an uncontrolled process that has the potential to affect plant operations. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. Escalation of this emergency classification level, if appropriate, would be based on HA3.1. PROCEDURE 5.7.1 REVISION 67

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 3 - Toxic, Corrosive, Asphyxiant, And Flammable Gas Initiating Condition: Access to a Vital Area is prohibited due to release of toxic, corrosive, asphyxiant, or flammable gases which jeopardizes operation of operable equipment required to maintain safe operations or safely shutdown the reactor EAL: HA3.1 Alert Access to any Table H-1 area is prohibited due to toxic, corrosive, asphyxlant, or flammable gases which jeopardize operation of systems required to maintain safe operations or safely shutdown the reactor (NOTE 7) NOTE 7 - If equipment in the stated area was already inoperable or out of service before the event occurred, then this EAL should not be declared as it wlll have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond _that_ already _allowed_ by Technical_ Specifications_ at_ the time of the _event. ___ _ I

                                                     ~  -- - -
                                                          '   -    -     L + , -~ I ~

Ta_ble H:.i Safe, Shutdow.rf~eas 1

                               '      -        O      T     -   ,-   *t :; -* - ', } -   ~ I

Reactor Building
Control Building
Service Water Pump Room
Diesel Generator Building
Cable Expansion Room NOTE - Examples of safety-related systems or components include: CRD (Control Rod Drive), CS (Core Spray), DG (Diesel Generator), and DG Support Systems, EE (Electrical Equipment, 4160 VAC, 480 VAC, 250 VDC, or 125 VDC), HPCI (High Pressure Coolant Injection), PC (Primary Containment), PCIS (Primary Containment Isolation System), REC (Reactor Equipment Cooling), RHR (Residual Heat Removal), RPS (Reactor Protection System), SGT (Standby Gas Treatment),

_SLC _(Standby_ Liquid Control), SW _(Service _Water).--------------------------------------------------- (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 165 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: All CNS Basis: This EAL is based on gases that have entered a plant structure in concentrations that could be unsafe for plant personnel and, therefore, preclude access to equipment necessary for the safe operation of the plant. Table H-1 safe shutdown areas contain systems.that are operated to establish or maintain safe shutdown (Reference 2-6). The Critical Switchgear Rooms are a part of the* Reactor Building. At CNS, there are various potential sources of atmospheric contamination. Some of these sources are:

Inert gas used for oxygen exclusion (nitrogen).
Combustion products.
Carbon dioxide from fire extinguishing.
Halon from fire extinguishing.
Welding gases (enclosed areas).
Vapors from painting (enclosed areas).
Vapors from petroleum products.
Hydrogen (OWC Hydrogen Gas Generation System, Generator Cooling System, batteries, and disassociation of water in the reactor).
Asphyxiants and irritants found most often in confined areas (water and oil storage tanks, open manholes).
Methane from bacterial action (tanks and pits).

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 166 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Some of the gases which could affect normal plant operations under this EAL are:

Carbon Monoxide - One of the most common asphyxlants encountered in industry. It is formed by the incomplete combustion of fuel containing carbon.

It may be found in the vicinity of a fire or a leak in an exhaust system (flue gas or internal combustion engines).

Oxygen - Oxygen has two fundamentally important properties: it supports combustion and it supports life. Since oxygen is necessary for life, it must be present in sufficient quantity. Oxygen deficiency occurs In confined spaces where the level of oxygen has been reduced below the limit to support life.

Oxygen content in the air can become fatally low in a brief period of time. Some of the more common causes of this problem are oxidation of metals, bacterial action, combustion, and displacement by other gases. An enriched oxygen atmosphere will accelerate combustion.

Hydrogen - Used in generator cooling. Hydrogen gas is also produced by the owe Gas Generation System, located in the owe Building, and subsequently injected into the Condensate System just upstream of the condensate booster pumps. Hydrogen is also produced by the disassociation of water from radiation in the reactor, which Is seen in the off-gas. The presence of hydrogen will be especially significant in the Off-Gas and Augmented Off-Gas Systems.

Hydrogen Is also a by-product of battery charging. It is lighter than air so it will be found in pockets at the ceiling of enclosures.

Argon - Commonly used during the welding of certain metals. It is denser than air so it will settle in pockets below the welding area.
Carbon Dioxide - Used to fight fire. Being heavier than air, carbon dioxide will settle in pockets and displace oxygen.
Halon - Used to fight fire in the Service Water Pump Room, Computer Room, and in the Simulator. Discharge of a Halon System will result in exceeding IDLH limits in the area. Discharge of the Service Water Pump Room or Computer Room Halon System should be classified under this EAL.
Nitrogen - Used primarily to purge Primary Containment. Since it is approximately the same density as air, it can be dispersed by proper ventilation. Areas of poor ventilation may contain greater than expected concentrations of nitrogen and consequently may be deficient in oxygen.
Combustible Gases and Vapors - Includes naturally occurring gases (such as methane and hydrogen gas) and the vapors of a large group of liquids which are used as fuels and solvents. Monitoring shall be required in fuel tanks .and other areas where explosive mixtures may be present.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 167 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

Hydrogen Sulfide - Classified as an irritant in low concentrations, but is even more toxic than carbon monoxide, because it inflames the mucus membranes and results in the lungs filling with fluid. This colorless gas has a characteristic rotten egg odor, which renders the sense of smell ineffective. Hydrogen sulfide may be found in sewage treatment or wherever organic matter containing sulfur decomposes and shall be monitored constantly during work.
Methane - The chief constituent of natural gas and is extremely explosive. It is non-toxic, but may reduce the oxygen content of an atmosphere, causing asphyxiation. Methane is often found in the vicinity of sanitary landfills and has beerJ detected in tanks where bacterial action is taking place (i.e., reactor water cleanup and condensate phase separator tanks). It is lighter than air and tends to accumulate in high spots or pockets. This can present a dangerous situation in storage tanks or sewers where access is normally gained at the top of the confined area.
Ethyl Benzene - Used primarily as an additive to diesel fuel. Acute exposure results in a local irritant effect on the skin and mucous membranes. Chronic exposure can lead to nervous system disorders and upper respiratory tract inflammation. Monitoring is required when entering a diesel fuel tank.
Chlorine - Used in chemical treatment of Circulate Water and Service Water Systems. Chlorine gas can be recognized by its pungent, irritating odor, which is like the odor of bleach. Chlorine is not flammable but can react explosively with other chemicals such as turpentine or ammonia. Chlorine gas stays close to the ground and spreads rapidly. When chlorine gas comes in contact with moist human tissues, such as the eyes throat and lungs, an acid is produced that can damage these tissues.
Chlorine Dioxide - This is a yellow to reddish-yellow manufactured gas which does not occur naturally in the environment. When added to water, chlorine dioxide forms chlorite ion, which is also a very reactive chemical. High levels of chlorine dioxide can be irritating to the nose, eyes, throat, and lungs.
Hydrogen Chloride - This is a colorless to slightly yellowish gas with a pungent odor. On exposure to air, the gas forms dense white vapors due to condensation with atmospheric moisture. The vapor is corrosive and air concentrations above 5 ppm can cause irritation. When mixed with water or atmospheric moisture, a highly corrosive atmosphere is formed. The most common source of Hydrogen Chloride gas is from Muriatic (Hydrochloric) Acid.

This EAL does not apply to routine inerting of the Primary Containment. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 168 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Administrative Procedure 0.36.6, Monitoring for Industrial Gases.

2. Site Services Procedure 1.1, Station Security.

3. CNS-FP-60, Fire Area Boundary Drawing Index.

4. Drawing CNS-EE-187, CNS Safe Shutdown Component Locations and Emergency Route Lighting - Site Plan.

5. Fire Safety Analysis Calculations.

6. USAR Section XII-2.1.2.1, Principal Class I Structures Required for Safe Shutdown.

NEI 99-01 Basis: Gases in a Safe Shutdown Area can affect the ability to safely operate or safely shutdown the reactor. The fact that SCBA may be worn does not eliminate the need to declare the event. Declaration should not be delayed for confirmation from atmospheric testing if the atmosphere poses an immediate threat to life and health or an immediate threat of severe exposure to gases. This could be based upon documented analysis, indication of personal ill effects from exposure, or operating experience with the hazards. If equipment in the stated area was already inoperable or out of service before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event. An asphyxiant Is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness, or even death. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 169 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either Operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL assumes concentrations of flammable gasses which can ignite/support combustion. Escalation of this emergency classification level, if appropriate, will be based on System Malfunctions, Fission Product Barrier Degradation, or Abnormal Rad Levels/Radioactive Effluent EALs. PROCEDURE 5. 7 .1 REVISION 67 PAGE 170 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Confirmed security condition or threat which indicates a potential degradation in the level of safety of the plant EAL: HU4.1 Unusual Event A security condition that does not involve a hostile action as reported by the Security Shift Supervisor OR A credible site-specific security threat notification OR A validated notification from NRC providing information of an aircraft threat Mode Applicability: All CNS Basis: Hostile Action: An act toward a nuclear power plant 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 i171cludes 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 CNS. 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). CNS Basis Reference(s):

1. CNS Security and Safeguards Contingency Plan.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 171 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: NOTE - Timely and accurate communication between Security Shift Supervision and_ the_ Control _Room _is crucial _for_ the_ implementation_ of_ effective Security EALs. __ _ Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10CFR73. 71 or in some cases under 10CFRS0. 72. Security events assessed as hostile actions are classifiable under HA4.1, HS4.1, and HG4.1. A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. Consideration should be given to upgrading the emergency response status and emergency classification level in accordance with the Physical Security Plan and Emergency Plan. 1st Threshold Reference is made to the Security Shift Supervisor because these Individuals are the designated personnel on-site 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 plant Physical Security Plan. This threshold is based on the CNS Security and Safeguards Contingency Plan. Security Plans are based on guidance provided by NE! 03-12. 2nd Threshold The second threshold is to ensure appropriate notifications for the security threat are made In a timely manner. This includes information of a credible threat. Only the plant to which the specific threat is made need to declare the Unusual Event. The determination of "credible" is made through use of information found in the Physical Security Plan. 3rd Threshold The Intent of this EAL threshold is to ensure notifications for the aircraft threat are made in a timely manner and that off-site response organizations and plant personnel are at a state of heightened awareness regarding the credible threat. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 172 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] This EAL is met when a plant receives information regarding an aircraft threat from NRC. Validation is performed by calling the NRC or by other approved methods of authentication. Only the plant to which the specific threat is made need to declare the Unusual Event. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an airliner (airliner is meant to be a large aircraft with the potential for causing significant damage to the plant). The status and size of the plane may be provided by NORAD through the NRC. Escalation to Alert emergency classification level via HA4.1 would be appropriate if the threat involves an airliner less than 30 minutes away from the plant. PROCEDURE 5.7.1 REVISION 67 PAGE 173 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Hostile action within the Owner Controlled Area or airborne attack threat© 1 EAL: HA4.1 Alert A hostile action is occurring or has occurred within the Owner Controlled Area as reported by the Security Shift Supervisor OR A validated notification from NRC of an airliner attack threat within 30 min. of the site Mode Applicability: All CNS Basis: Reference is made to the Security Shift Supervisor because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred. Hostile Action: An act toward a nuclear power plant 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 deUver 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 CNS. 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). CNS Basis Reference(s):

1. CNS Security and Safeguards Contingency Plan.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 174 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: NOTE - Timely and accurate communication between Security Shift Supervision _and_ the_ Control_ Room _is crucial _for_ the_ implementation _of_ effective Security EALs. __ _ These EAL thresholds address the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. They are not premised solely on the potential for a radiological release. Rather, the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land, or water attack elements. The fact that the site is under serious attack or is an identified attack target with minimal time available for further preparation or additional assistance to arrive requires a heightened state of readiness and Implementation of protective measures that can be effective (such as on-site evacuation, dispersal, or sheltering). 1st Threshold This EAL threshold addresses the potential for a very rapid progression of events due to a hostile action. It is.not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the Owner Controlled Area. Those events are adequately addressed by other EALs. Although nuclear plant Security Officers are well trained and prepared to protect against hostile action, i_t is appropriate for off-site response organizations to be notified and encouraged to begin activation (if they do not normally) to be better prepared should it be necessary to consider further actions. 2nd Threshold This EAL threshold addresses the immediacy of an expected threat arrival or impact on the site within a relatively short time. The intent of this EAL threshold is to ensure notifications for the airliner attack threat are made in a timely manner and that off-site response organizations and plant personnel are at a state of heightened awareness regarding the credible threat. Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 175 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] This EAL threshold is met when a plant receives information regarding an airliner attack threat from NRC and the airliner is within 30 minutes of the plant. Validation is performed by calling the NRC or by other approved methods of authentication. Only the plant to which the specific threat is made need to declare the Alert. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an airliner. The status and size of the plane may be provided by NORAD through the NRC. If not previously notified by the NRC that the airborne hostile action was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA, or NRC. However, the declaration should not be unduly delayed awaiting Federal notification. PROCEDURE 5. 7 .1 REVISION 67 PAGE 176 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Hostile action within the Protected Area© 1 EAL: HS4.1 Site Area Emergency A hostile action is occurring or has occurred within the Protected Area as reported by the Security Shift Supervisor Mode Applicability: All CNS Basis: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs. Reference is made to the Security Shift Supervisor because this individual is the designated on-site person 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 CNS Security and Safeguards Contingency Plan (Reference 1). Hostile Action: An act toward a nuclear power plant 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 CNS. 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). CNS Basis Reference(s):

1. CNS Security and Safeguards Contingency Plan.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 177 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This condition represents an escalated threat to plant safety above that contained in the Alert in that a hostile force has progressed from the Owner -controlled Area to the Protected Area. This EAL addresses the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. It is not premised solely on the potential for a radiological release. Rather, the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land, or water attack elements. The fact that the site is under serious attack with minimal time available for further preparation or additional assistance to arrive requires off-site response organizations readiness and preparation for the implementation of protective measures. This EAL addresses the potential for a very rapid progression of events due to a hostile action. It is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the Protected Area. Although nuclear plant Security Officers are well trained and prepared to protect against hostile action, it ls appropriate for off-site response organizations to be notified and encouraged to begin preparations for public protective actions (if they do not normally) to be better prepared should it be necessary to consider further actions. If not previously notified by NRC that the airborne hostile action was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA, or NRC. However, the declaration should not be unduly delayed awaiting Federal notification. Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack. PROCEDURE 5. 7 .1 REVISION 67 PAGE 178 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 4 - Security Initiating Condition: Hostile action resulting in loss of physical control of the facility© 1 EAL: HG4.1 General Emergency A hostile action has occurred such that plant personnel are unable to operate equipment required to maintain safety functions OR A hostile action has caused failure of Spent Fuel Cooling Systems and imminent fuel damage is likely for a freshly off-loaded reactor core in pool Mode Applicability: All CNS Basis: A freshly off-loaded reactor core in pool consists of recently discharged fuel that has been out of the reactor for less than 1 year (Reference 1). CNS Basis Reference(s):

1. Nuclear Procedure 10.6, SFSP Fuel Storage Constraints (restricted access for B.5.b).

2. CNS Security and Safeguards Contingency Plan.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 179 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: 1st Threshold This EAL threshold encompasses conditions under which a hostile action has resulted in a loss of physical control of Vital Areas (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. These safety functions are reactivity control (ability to shut down the reactor and keep it shut down), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink). Loss of physical control of the Control Room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the threshold ls not met. 2nd Threshold

This EAL threshold addresses failure of Spent Fuel Cooling Systems as a result of hostile action If imminent fuel damage Is likely, such as when a freshly off-loaded reactor core is in the spent fuel pool.

PROCEDURE 5. 7.1 REVISION 67 PAGE 180 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 5 - Control Room Evacuation Initiating Condition: Control Room evacuation has been initiated EAL: HAS.1 Alert Procedure 5.lASD, Alternate Shutdown, or Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room, requires Control Room evacuation Mode Applicability: All CNS Basis: Procedures 5.lASD, Alternate Shutdown (Reference 1), and 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room (Reference 2), provide the instructions for scramming the unit and maintaining RCS inventory from outside the Control Room. The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. Control Room inhabitabillty may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. CNS Basis Reference(s):

1. Emergency Procedure 5.lASD, Alternate Shutdown.

2. Emergency Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room.

NEI 99-01 Basis: With the Control Room evacuated, additional support, monitoring, and direction through the Technical Support Center and/or other Emergency Response Facilities may be necessary. Inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency. PROCEDURE 5.7.1 REVISION 67 PAGE 181 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 5 - Control Room Evacuation Initiating Condition: Control Room evacuation has been Initiated and plant control cannot be established EAL: HS5.1 Site Area Emergency Control Room evacuation has been initiated AND Control of the plant cannot be established within 15 min Mode Applicability: All CNS Basis: Procedures 5. lASD, Alternate Shutdown (Reference 1), and 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room (Reference 2), provide the instructions for scramming the unit and maintaining RCS inventory from outside the Control Room. 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. The 15 minute criterion applies from the time the Control Room begins to be evacuated. CNS Basis Reference(s):

1. Emergency Procedure 5.lASD, Alternate Shutdown.

2. Emergency Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room.

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 182 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The intent of this EAL is to capture those events where control of the plant cannot be re-established in a timely manner. In this case, expeditious transfer of control of safety systems has not occurred (although fission product barrier damage may not yet be indicated). The 15 minute time for transfer starts when the Control Room begins to be evacuated (not when Procedure 5. lASD, Alternate Shutdown, is entered). The time interval is based on how quickly control must be re-established without core uncovery and/or core damage. The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shut down the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink). The determination of whether or not control is established from outside the Control Room is based on Emergency Director (ED) judgment. The Emergency Director is expected to make a reasonable, informed judgment that control of the plant from the Alternate Shutdown Panels cannot be established within the 15 minute interval. Escalation of this emergency classification level, if appropriate, would be by Fission Product Barrier Degradation or Abnormal Rad Levels/Radiological Effluent EALs. PROCEDURE 5.7.1 REVISION 67 PAGE 183 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a UE EAL: HU6.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director indicate that EITHER: Events are in-progress or have occurred which indicate a potential degradation of the level of safety of the plant OR A security threat to facility protection has been initiated No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs Mode Applicability: All CNS Basis: The Emergency Director is the designated on-site individual having the responsibility and authority for implementing the CNS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency 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 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, 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 (Reference 1). (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 184 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. Emergency Plan for Cooper Nuclear Station, Section 5.0.

NEI 99-01 Basis: This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Unusual Event emergency class. PROCEDURE 5.7.1 REVISION 67 PAGE 185 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H ~ Hazards And Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert EAL: HA6.1 Alert Other conditions exist which in the judgment of the Emergency Director indicate that events are in-progress or have occurred which Involve EITHER: 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 beyond the site boundary Mode Applicability: All CNS Basis: The Emergency Director is the designated on-site individual having the responsibility and authority for implementing the CNS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency 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 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, 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 (Reference 2). I For the purposes of this EAL, the Site Boundary for CNS Is a 1 mile radius around the plant. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 186 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. CNS Drawing DWG.2.2 (P3-A-45).

2. Emergency Plan for Cooper Nuclear Station, Section 5.0.

NEI 99-01 Basis: This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency class. PROCEDURE 5. 7 .1 REVISION 67 PAGE 187 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of Site Area Emergency EAL: HS6.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in-progress or have occurred which involve EITHER: An actual or likely major failures of plant functions needed for protection of the public OR Hostile action that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of; or 2) that prevent effective access to equipment needed for the protection of the public Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels (1 Rem TEDE and 5 Rem thyroid CDE) beyond the site boundary Mode Applicability: All ( continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 188 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: The Emergency Director is the designated on-site individual having the responsibility and authority for Implementing the CNS Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan Implementing procedures. If required by the emergency classlflcation or if deemed appropriate by the Emergency 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, 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 (Reference 2). For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant. CNS Basis Reference(s):

1. CNS Drawing DWG.2.2 (P3-A-45).

2. Emergency Plan for Cooper Nuclear Station, Section 5.0.

NEI 99-01 Basis: This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency class description for Site Area Emergency. PROCEDURE 5. 7 .1 REVISION 67 PAGE 189 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: H - Hazards And Other Conditions Affecting Plant Safety Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of General Emergency EAL: HG6.1 General Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in-progress or have occurred which involve EITHER: 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 (1 Rem TEDE and 5 Rem thyroid COE) beyond the site boundary Mode Applicability: All CNS Basis: The Emergency Director ls the designated on-site individual having the responsibility and authority for implementing the CNS Emergency Plan. The Shift Manager (SM) Initially acts in the capacity of the Emergency 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 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, 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 (Reference 2). (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 190 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary. For the purposes of this EAL, the Site Boundary for CNS is a 1 mile radius around the plant. CNS Basis Reference(s):

1. CNS Drawing DWG.2.2 (P3-A-45).

2. Emergency Plan for Cooper Nuclear Station, Section 5.0.

NEI 99-01 Basis: This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the General Emergency class. PROCEDURE 5. 7 .1 REVISION 67 PAGE 191 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: Loss of all off-site AC power to critical buses for 15 minutes or longer EAL: SUl.1 Unusual Event Loss of all off-site AC power (Table S-3) to critical 4160V Buses lF and lG for

   ~  15 min. (NOTE 3)

NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition has exceeded or_ wi 11 _Ii kely_ exceed_ the a ppl ica ble_ ti me*------------------------------------------------------ Table 5-3 AC Power Sources Offsite

Startup Station Service Transformer
Emergency Station Service Transformer
Backfeed 345 kv line through Main Power Transformer to the Normal Station Service Transformer (Note 8)

Onsite

DG-1
DG-2
Main Generator r------------------------------------------------------------------------------------------------------------------------------------

NOTE 8 - The time required to establish the backfeed is likely longer than the

! specified time interval. If off-normal plant conditions have already established the ! backfeed, its power to the safety-related buses may be considered an off-site

' power source.

L------------------------------------------------------------------------------------------------------------------------------------- (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 192 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: The 4160V critical Buses 1F (Div. I) and 1G (Div. II) are the plant essential, safety-related emergency buses. Each can be energized manually and separately by any of the following off-site sources of power: Figure S-1 illustrates the 4160V AC Distribution System (Reference 7, 8).

Startup Transformer - The Startup Transformer provides a source of off-s1te AC power to the entire Auxiliary Power Distribution System adequate for the startup operation or shutdown operation of the station. The Startup Transformer is the preferred source of off-site AC power to the station whenever the main generator is off-line ( < 160 MWe). The Startup Transformer is energized from the 161 kV Switchyard. The transformer is normally left energized at all times to provide for quick automatic transfer of the 4160V auxiliaries to the Startup Transformer in the event the station Normal Transformer fails or main generator trips off-line.
Emergency Transformer - The Emergency Transformer is the primary off-site AC power source to essential station loads. During normal station operation, the Emergency Transformer is energized by the 69 kV transmission line from OPPD. As such, it supplies 4160V Switchgear 1F and/or 1G in the event the Normal Transformer and Startup Transformer are not available for service. Use of the Emergency Transformer also allows portions of the 345 kV System to be removed from service for inspection, testing, and maintenance.
Backfeeding power from the 345 kV line through the Main Power Transformer to the Normal Transformer. The Normal Transformer is the normal source of AC power to the station when the Main Generator is on line above 20%

(160 MWe) electrical power. The transformer is energized during Main Generator operation through the Isolated Phase Buses that feed the Main Power Transformers. As mentioned In NOTE 8, the time required to establish the backfeed is likely longer than the 15 minute interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site power source. On-site power sources are the emergency diesel generators (DG-1 and DG-2) and the Main Generator. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 193 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If a power supply should have picked up a critical bus but failed to do so, that power supply should be considered unavailable until it has been successfully tied onto the bus. If power supply can be tied to the bus within 15 minutes, then the power supply is to be considered available. The Supplemental Diesel Generator (SDG) is not considered an on-site or an off-site emergency power supply and is not considered in classifications involving loss of power. The SBO Coping Time per Regulatory Guide 1.155 considers the impact of a SDG. The 15 minute interval was selected as a threshold to exclude transient or momentary power losses. If neither emergency bus is energized by an off-site source within 15 minutes, an Unusual Event is declared under this EAL. CNS Basis Reference(s):

1. System Operating Procedure 2.2.15, Startup Transformer.

2. System Operating Procedure 2.2.16, Normal Station Service Transformer.

3. System Operating Procedure 2.2.17, Emergency Station Service Transformer.

4. System Operating Procedure 2.2.18.1, 4160V Auxiliary Power Distribution System.

5. System Operating Procedure 2.2.18.3.DIVl, 4160V Div 1 Distribution Support.

6. System Operating Procedure 2.2.18.3.DIV2, 4160V Div 2 Distribution Support.

7. System Operating Procedure 2.2.18.4, 4160V Distribution Abnormal Power.

8. System Operating Procedure 2.2.20, Standby AC Power System (Diesel Generator).

9. Emergency Procedure 5.3580, Station Blackout.

10. BR 3001, One Line Diagram.

11. BR 3002, Sheet 1.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 194 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

12. NC43456, One Line Switching Diagram 161kV Substation.

13. Enercon Services, Inc. Report No. NPPl-PR-01, Station Blackout Coping Assessment for Cooper Nuclear Station, Revision 2.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 195 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure S-1: 4160V AC Distribution System FROM FROM MAIN GENERATOR 345 KVN61 KV GRID y NORMAL STARTUP y V,J.J,J STATION SERVICE

          ~        STATION SERVICE                                           TRANSFORMER        rvYV'\

TRANSFORMER 161 KV/4160V 22 KV/4160\/ 4180\/ SWGR 1C 4160V SWGR 10 BKR BKR )BKR 1AN ) 1BN 1BS 4160V SWGR 1A

                                                                ')BKR *           ')BKR

4160V SWGR 1B 1BE 1BG 4160V SWGR 1E BKR 1FA 4160V SWGR 1F 4160V SWGR 1G BKR 1FE

                  ')              BKR 1FS' )                                   ')BKR 1GS          ')BKR1GE I                                     I MODS BKR EG1
                    )

I MOOO:.RG~ J:¢¢¢:CsTATION SERVICE )~ I c5 c5 TRANSFORMER SOG DIESEL GENERATOR #1 DIESEL GENERATOR #2 (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 196 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (e.g., Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. PROCEDURE 5.7.1 REVISION 67 PAGE 197 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: AC power capability to critical buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in loss of all AC power to critical buses EAL: SA1.1 Alert AC power capability to critical 4160V Buses 1F and 1G reduced to a single power source (Table 5-3) for~ 15 min. (NOTE 3) such that any additional single failure would result in loss of all AC power to emergency buses NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will _Ii kely _exceed_ the_ a p pl ica ble _time. _______________________________________________________________________________ _ Table 5-3 AC Power Sources Offslte

Startup Station Service Transformer
Emergency Station Service Transformer
Backfeed 345 kv line through Main Power Transformer to the Normal Station Service Transformer (Note 8)

Onslte

DG-1
DG-2
Main Generator NOTE 8 - The time required to establish the backfeed is likely longer than the specified time interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site

_power source. _____________________________________________________________________________________________________________ _ (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 198 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: The 4160V critical Buses lF (Div. I) and lG (Div. II} are the plant essential, safety-related emergency buses. Each can be energized manually and separately by any of the following off-site sources of power: Figure 5-1 illustrates the 4160V AC Distribution System (Reference 7, 8).

Startup Transformer - The Startup Transformer provides a source of off-site AC power to the entire Auxiliary Power Distribution System adequate for the startup operation or shutdown operation of the station. The Startup Transformer is the preferred source of off-site AC power to the station whenever the main generator Is off-line ( < 160 MWe). The Startup Transformer is energized from the 161 kV Switchyard. The transformer is normally left energized at all times to provide for quick automatic transfer of the 4160V auxiliaries to the Startup Transformer in the event the station Normal Transformer fails or main generator trips off-line.
Emergency Transformer - The Emergency Transformer is the primary off-site AC power source to essential station loads. During normal station operation, the Emergency Transformer is energized by the 69 kV transmission line from OPPD. As such, it supplies 4160V Switchgear lF and/or lG in the event the Normal Transformer and Startup Transformer are not available for service. Use of the Emergency Transformer also allows portions of the 345 kV System to be removed from service for inspection, testing, and maintenance.
Backfeeding power from the 345 kV line through the Main Power Transformer to the Normal Transformer. The Normal Transformer is the normal source of AC power to the station when the Main Generator is on-line above 20%

(160 MWe) electrical power. The transformer is energized during Main Generator operation through the Isolated Phase Buses that feed the Main Power Transformers. As mentioned in NOTE 8, the time required to establish the backfeed is likely longer than the 15 minute interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site power source. On-site power sources are the emergency diesel generators (DG-1 and DG-2) and the Main Generator. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 199 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATIO-N USE] If a power supply should have picked up a critical bus but failed to do so, that power supply should be considered unavailable until it has been successfully tied onto the bus. If power supply can be tied to the bus within 15 minutes, then the power supply is to be considered available. The Supplemental Diesel Generator (SDG) is not considered an on-site or an off-site emergency power supply and is not considered in classifications involving loss of power. The SBO Coping Time per Regulatory Guide 1.155 considers the impact of a SDG. 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. CNS Basis Reference(s):

1. System Operating Procedure 2.2.15, Startup Transformer.

2. System Operating Procedure 2.2.16, Normal Station Service Transformer.

3. System Operating Procedure 2.2.17, Emergency Station Service Transformer.

4. System Operating Procedure 2.2.18.1, 4160V Auxiliary Power Distribution System.

5. System Operating Procedure 2.2.18.3.DIVl, 4160V Div 1 Distribution Support.

6. System Operating Procedure 2.2.18.3.DIV2, 4160V Div 2 Distribution Support.

7. System Operating Procedure 2.2.18.4, 4160V Distribution Abnormal Power.

8. System Operating Procedure 2.2.20, Standby AC Power System (Diesel Generator).

9. Emergency Procedure 5.3SBO, Station Blackout.

10. BR 3001, One Line Diagram.

11. BR 3002, Sheet 1.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 200 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

12. NC43456, One Line Switching Diagram 161kV Substation.

13. Enercon Services, Inc. Report No. NPPl-PR-01, Station Blackout Coping Assessment for Cooper Nuclear Station, Revision 2.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 201 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure S-1: 4160V AC Distribution System FROM FROM MAIN GENERATOR 345 KV/161 KV GR10 y N~ STARTUP y

                                                                                            \.AAA.I STATION SERVICE
           ~       STATION SERVICE                                        TRANSFORMER       rvvY\

TRANSFORMER 161 KV/4160V 22KV/4160V

                   ~)
                       ~ 4160V SWGR 1C                            4160V SWGR 1D BKR                                                  BKR                )BKR 1AN                                                  1BN                   1BS
                                                                      )

4160VS'NGR1A

                                                             ')BKR            ')BKR

4160VSWGR 18 1BE 1BG 4160V SWGR 1E BKR 1FA 4160V SWGR 1F 4160V SWGR 1G BKR ')

1FE BKR') 1FS ')BKR1GS ')BKR1GE I I

                                           ~:.~~

MODS BKR EG1

                     )                   I
                                      ~STATION SERVICE                                      )

BKR EG2 I c5 c5 TRANSFORMER SOG DIESEL GENERATOR #1 DIESEL GENERATOR #2. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 202 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is intended to provide an escalation from IC SU1.1, "Loss of all off-site AC power to critical buses for 15 minutes or longer". The condition indicated by this EAL is the degradation of the off-site and on-site AC power systems such that any additional single failure would result in a loss of all AC power to the critical buses. This condition could occur due to a loss of off-site power with a concurrent failure of all but one emergency generator to supply power to its critical bus. Another related condition could be the loss of all off-site power and loss of on-site emergency generators with only one train of crltical buses being backfed from the unit main generator or the loss of on-site emergency generators with only one train of critical buses being backfed from off-site power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with 551.1. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. PROCEDURE 5.7.1 REVISION 67 PAGE 203 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: Loss of all off-site and all on-site AC power to critical buses for 15 minutes or longer EAL: SSl.1 Site Area Emergency Loss of all off-site and all on-site AC power (Table S-3) to critical 4160V Buses lF and lG for c: 15 min. (NOTE 3) NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will Ii kely _exceed_ the_ a pp Ii cab Ie _time. _______________________________________________________________________________ _ Table S-3 AC Power Sources Offsite

Startup Station Service Transformer
Emergency Station Service Transformer
Backfeed 345 kv line through Main Power Transformer to the Normal Station Service Transformer (Note 8)

Onsite

DG-1
DG-2 '
Main Generator NOTE 8 - The time required to establish the backfeed is likely longer than the specified time interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site

_power source. _____________________________________________________________________________________________________________ _ (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 204 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: The 4160V critical Buses lF (Div. I) and lG (Div. II) are the plant essential, safety-related emergency buses. Each can be energized manually and separately by any of the following off-site sources of power: Figure S-1 illustrates the 4160V AC Distribution System (Reference 7, 8).

Startup Transformer - The Startup Transformer provides a source of off-site AC power to the entire Auxiliary Power Distribution System adequate for the startup operation or shutdown operation of the station. The Startup Transformer is the preferred source of off-site AC power to the station whenever the main generator is off-line ( < 160 MWe). The Startup Transformer is energized from the 161 kV Swltchyard. The transformer is normally left energized at all times to provide for quick automatic transfer of the 4160V auxiliaries to the Startup Transformer in the event the station Normal Transformer fails or main generator trips off-line.
Emergency Transformer - The Emergency Transformer is the primary off-site AC power source to essential station loads. During normal station operation, the Emergency Transformer is energized by the 69 kV transmission line from OPPD. As such, it supplies 4160V Switchgear lF and/or lG in the event the Normal Transformer and Startup Transformer are not available for service. Use of the Emergency Transformer also allows portions of the 345 kV System to be removed from service for inspection, testing, and maintenance.
Backfeeding power from the 345 kV line through the Main Power Transformer to the Normal Transformer. The Normal Transformer is the normal source of AC power to the station when the Main Generator is on line above 20%

(160 MWe) electrical power. The transformer is energized during Main Generator operation through the Isolated Phase Buses that fee_d the Main Power Transformers. As mentioned in NOTE 8, the time required to establish the backfeed is likely longer than the 15 minute interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site power source. On-site power sources are the emergency diesel generators (DG-1 and DG-2) and the Main Generator. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 205 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If a power supply should have picked up a critical bus but failed to do so, that power supply should be considered unavailable until it has been successfully tied onto the bus. If power supply can be tied to the bus within 15 minutes, then the power supply is to be considered available. The Supplemental Diesel Generator (SDG) is not considered an on-site or an off-site emergency power supply and is not considered in classifications involving loss of power. The SBO Coping Time per Regulatory Guide 1.155 considers the impact of a SDG. This EAL is the hot condition equivalent of the cold condition loss of all AC power EAL CAL 1. When in Cold Shutdown, Refueling, or Defueled Mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the critical buses, relative to that existing when in hot conditions. CNS Basis Reference(s):

1. System Operating Procedure 2.2.15, Startup Transformer.

2. System Operating Procedure 2.2.16, Normal Station Service Transformer.

3. System Operating Procedure 2.2.17, Emergency Station Service Transformer.

4. System Operating Procedure 2.2.18.1, 4160V Auxiliary Power Distribution System.

5. System Operating Procedure 2.2.18.3.DIVl, 4160V Div 1 Distribution Support.

6. System Operating Procedure 2.2.18.3.DIV2, 4160V Div 2 Distribution Support.

7. System Operating Procedure 2.2.18.4, 4160V Auxiliary Power Distribution System.

8. System Operating Procedure 2.2.20, Standby AC Power System (Diesel Generator).

9. Emergency Procedure 5.3SBO, Station ~lackout.

10. BR 3001, One Line Diagram.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 206 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

11. BR 3002, Sheet 1.

12. NC43456, One Line Switching Diagram 161kV Substation.

13. Enercon Services, Inc. Report No. NPPl-PR-01, Station Blackout Coping Assessment for Cooper Nuclear Station, Revision 2.

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 207 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure S-1: 4160V AC Distribution System FROM FROM MAIN GENERATOR 345 KV/161 KV GRID y ~ STARTUP y

                                                                                                  \AA.A.I STATION SERVICE
            ~        STATION SERVlCE                                            TRANSFORMER       rvvY'\

TRANSFORMER 161 KV/4160V 22 KV/4160V BKR 1CN ) BKR 1CS )

                                                                       )~                 )~
                         ~4160VSWGR 1C         *
                                                                     * *4160V SWGR 1D 4160V SWGR 1A
                                                                   ')BKR 1BE
                                                                                    ')BKR1BG 4160V SWGR 1B 4160V SWGR 1E
                                                                   '~

BKR 1FA )~ 4160V SWGR 1F BKR 1FE ' ) Bl<R 1FS' ) ')BKR l 1GS 4160V SWGR 1G

                                                                                                ')BKR1GE I                 1                    I
                                              '-=,

MOOS BKR EG1

                      )                  ~STATION SERVICE                                         )~

I c5 c5 TRANSFORMER SDG DIESEL GENERATOR #1 DIESEL GENERATOR #2 (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 208 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Loss of all AC power to critical buses compromises all plant safety systems requiring electric power including RHR, ECCS, containment heat removal, and the ultimate heat sink. Prolonged loss of all AC power to critical buses will lead to loss of Fuel Clad, RCS, and Primary Containment, thus, this event can escalate to a General Emergency. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site power. Escalation to General Emergency is via Fission Product Barrier Degradation or IC SG1, "Prolonged loss of all off-site power and prolonged loss of all on-site AC power". PROCEDURE 5.7.1 REVISION 67 PAGE 209 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: Prolonged loss of all off-site and all on-site AC power to critical emergency buses EAL: SGl.1 General Emergency Loss of all off-site and all on-site AC power (Table 5-3) to critical 4160V Buses lF and lG, AND EITHER: Restoration of at least one emergency bus in < 4 hours is not likely OR RPV level cannot be restored and maintained > -158 in. or cannot be determined Table 5-3 AC Power Sources Offslte

Startup Station Service Transformer
Emergency Station Service Transformer
Backfeed 345 kv line through Main Power Transformer to the Normal Station Service Transformer (Note 8)

Onsite

DG-1
DG-2
Main Generator NOTE 8 - The time required to establish the backfeed is likely longer than the specified time interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be considered an off-site

_power source. _____________________________________________________________________________________________________________ _ (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 210 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: The 4160V Critical Buses 1F (Div. I) and 1G (Div. II) are the plant essential, safety-related emergency buses. Each can be energized manually and separately by any of the following off-site sources of power: Figure 5-1 illustrates the 4160V AC Distribution System (Reference 7, 8).

Startup Transformer - The Startup Transformer provides a source of off-site AC power to the entire Auxiliary Power Distribution System adequate for the startup operation or shutdown operation of the station. The Startup Transformer is the preferred source of off-site AC power to the station whenever the main generator is off-line ( < 160 MWe). The Startup Transformer is energized from the 161 kV Switchyard. The transformer is normally left energized at all times to provide for quick automatic transfer of the 4160V auxiliaries to the Startup Transformer in the event the station Normal Transformer fails or main generator trips off-line.
Emergency Transformer - The Emergency Transformer is the primary off-site AC power source to essential station loads. During normal station operation, the Emergency Transformer is energized by the 69 kV transmission line from OPPD. As such, it supplies 4160V Switchgear 1F and/or 1G in the event the Normal Transformer and Startup Transformer are not available for service. Use of the Emergency Transformer also allows portions of the 345 kV System to be 1

removed from service for inspection, testing, and maintenance.

Backfeeding power from the 345 kV line through the Main Power Transformer to the Normal Transformer. The Normal Transformer is the normal source of AC power to the station when the Main Generator is on line above 20%

(160 MWe) electrical power. The transformer is energized during Main Generator operation through the Isolated Phase Buses that feed the Main Power Transformers. As mentioned in NOTE 8, the time required to establish the backfeed is likely longer than the 4 hour interval. If off-normal plant conditions have already established the backfeed; however, its power to the safety-related buses may be considered an off-site power source. On-site power sources are the emergency diesel generators (DG-1 and DG-2) and the Main Generator.

                       ..J (continued on next page)

PROCEDURE 5.7.1 REVISION 67 PAGE 211 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If a power supply should have picked up a critical bus but failed to do so, that power supply should be considered unavailable until it has been successfully tied onto the bus. 4 hours is the CNS Station Blackout Coping Analysis time (Reference 11, 13). The Supplemental Diesel Generator (SDG) is not considered an on-site or an off-site emergency power supply and is not considered in classifications involving loss of power. The SBO Coping Time per Regulatory Guide 1.155 considers the impact of a SDG. Indication of continuing core cooling degradation is manifested by a RPV level instrument reading of< -158 inches (RPV level is below the top of active fuel). When RPV level is at or above the top of active fuel, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below the top of active fuel, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to the top of active fuel, the level is indicative of a challenge to core cooling and the Fuel Clad barrier. When RPV level cannot be determined, EOPs require entry to EOP-2B, RPV Flooding, or EOP-7B, RPV Flooding (Failure-to-Scram). RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-2B/7B specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 212 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

; . System Operating Procedure 2.2.15, Startup Transformer.

2. System Operating Procedure 2.2.16, Normal Station Service Transformer.

3. System Operating Procedure 2.2.17, Emergency Station Service Transformer.

4. System Operating Procedure 2.2.18.1, 4160V Auxiliary Power Distribution System.

5. System Operating Procedure 2.2.18.3.DIVl, 4160V Div 1 Distribution Support.

6. System Operating Procedure 2.2.18.3.DIV2, 4160V Div 2 Distribution Support.

7. System Operating Procedure 2.2.18.4, 4160V Distribution Abnormal Power.

8. System Operating Procedure 2.2.20, Standby AC Power System (Diesel Generator).

9. Emergency Procedure 5.3SBO, Station Blackout.

10. BR 3001, One Line Diagram.

11. BR 3002, Sheet 1.

12. NC43456, One Line Switching Diagram 161kV Substation.

13. EOP-2B RPV Flooding.

14. EOP-7B RPV Flooding (Failure-to-Scram).

15. Enercon Services, Inc. Report No. NPP1-PR-01, Station Blackout Coping Assessment for Cooper Nuclear Station, Revision 2.

16. NEDC 97-089.

17. USAR Section VIll-6.2.7.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 213 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure S-1: 4160V AC Distribution System FROM FROM MAIN GENERATOR 346 KV/161 KV GRID y NORMAL STARTUP y

                                                                                            \AAA.I STATION SERVICE
           ~       STATION SERVICE                                        TRANSFORMER       fV"(V'\

TRANSFORMER 161 KV/4160V 22 KV/4160V

                                                                 )~                )~

4160V SWGR tC

                                                               ~  4160V SWGR 1D
                                                                                 ~

4160V SWGR 1A

                                                             ')BKR            ')BKR       416(]1/ SWGR  1B 1BE               1BG 4160V SWGR 1E r

BKR 1FA 4160V SWGR 1F 4160\J SWGR 1G BKR 1FE ' ) BKR') 1FS ')BKR1GS ')BKR 1GE I I

               ~)                       I ~~-G:
                                      ~STATION SERVICE I

II/ODS

BKR EG2 c5 c5 TRANSFORMER SDG DIESEL GENERATOR #1 DIESEL GENERATOR #2 (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 214 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Loss of all AC power to critical buses compromises all plant safety systems requiring electric power including RHR, ECCS, containment heat removal, and the ultimate heat sink. Prolonged loss of all AC power to critical buses will lead to loss of fuel clad, RCS, and containment, thus, warranting declaration of a General Emergency. This EAL is specified to assure that in the unlikely event of a prolonged loss of all critical bus AC power, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as ls appropriate, based on a reasonable assessment of the event trajectory. The likelihood of restoring at least one critical bus should be based on a realistic appraisal of the situation since a delay in .an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions. In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of fission product barriers is imminent?

2. If there are no present Indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?

Thus, indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers. PROCEDURE 5. 7 .1 REVISION 67 PAGE 215 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Inadvertent criticality EAL: SU2.1 Unusual Event An unplanned sustained positive period observed on nuclear instrumentation Mode Applicability: 3 - Hot Shutdown CNS Basis: SRM A-D period Meters NMS-I-44A-D on Panel 9-5 identify this condition as well as Panel 9-5 amber light and SRM Period ( < 50 sec.) Annunciator 9-5-1/F-8 (Reference 1, 2). However, a SRM period alarm caused by SRM channel noise does not result in entry into this EAL (Reference 1). CNS Basis Reference(s):

1. Alarm Procedure 2.3_9-5-1, Panel 9 Annunciator 9-5-1, F-8, SRM Period.

2. Instrument Operating Procedure 4.1.1, Source Range Monitoring System.

NEI 99-01 Basis: This EAL addresses inadvertent criticality events. This EAL indicates a potential degradation of the level of safety of the plant, warranting an Unusual Event classification. This EAL excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups ( e.g., criticality earlier than estimated). Escalation would be by the Fission Product Barrier Table, as appropriate, to the operating mode at the time of the event. PROCEDURE 5. 7.1 REVISION 67 PAGE 216 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Automatic scram fails to shut down the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor EAL: SA2 .. 1 Alert An automatic scram failed to shut down the reactor AND Manual actions taken at the reactor control console (NOTE 5) successfully shut down the reactor as indicated by reactor power < 3% NOTE 5 - Manual scram methods for EAL SA2.1 and EAL 552.1 are the following:

Reactor Scram pushbuttons.
Reactor Mode switch in SHUTDOWN.
Manual or auto actuation of ARI.

Mode Applicability: 1 - Power Operation, 2 - Startup CNS Basis: The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints. A reactor scram may be the result of manual or automatic action in response to any of the following parameters (Reference 3):

APRM Fixed Neutron Flux - High ..
APRM Fixed Neutron Flux - High (Setdown).
APRM Flow Biased - High.
IRM - High.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 217 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

Reactor Vessel Water Level Low - Level-3.
Turbine Stop Valve Closure.
Turbine Control Valve Fast Closure.
MSIV Closure.
Scram Discharge Volume (SDV) Level - High.
Drywell Pressure - High.

Following a successful reactor scram, 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 period. 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-scram response from an automatic reactor scram 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 scram has therefore occurred when there is sufficient rod insertion to bring the reactor power below the APRM downscale setpoint of 3% (Reference 1, 2). The significance of the second condition of this EAL is that a potential degradation of a safety system exists because a front line automatic protection system did not function in response to a plant transient. Thus, plant safety has been compromised. Following any automatic RPS scram signal, Procedure 2.1.5_ prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown ls achieved (Reference 1). Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Alert. This EAL is not applicable if a manual scram is initiated and no RPS setpoints are exceeded. Taking the mode switch to shutdown is a manual scram action. When the mode switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 218 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] In the event the Operator identifies a reactor scram Is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. Methods of inserting a manual scram are limited to those that can be taken rapidly at the reactor control console (Panel 9-5) and include:

Both manual Reactor Scram pushbuttons.
Reactor Mode switch in SHUTDOWN.
Manual or auto actuation of ARI.

Auto actuation of ARI is included in the list of methods because the Operator, by procedure, always ensures actuation of ARI has occurred if the ARI actuation setpoints are exceeded. This means action to depress the ARI pushbuttons is taken if the automatic ARI actuation setpoints are exceeded but failed to actuate. If ARI properly actuates automatically, the ARI pushbuttons are not depressed. Reactor shutdown achieved by use of the alternate rod insertion methods listed in Procedure 5.8.3 do not constitute a successful manual scram (Reference 2). The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power below 3% (Reference 1, 2), the event escalates to the Site Area Emergency under EAL 552.1. If by procedure, Operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals the automatic scram did not caus'e 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. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 219 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. General Operating Procedure 2.1.5, Reactor Scram.

2. Emergency Procedure 5.8.3, Alternate Rod Insertion Methods.

3. AMP-TBD00, Appendix B, Step RC/Q-4.

4. USAR Table VII-2-2.

NEI 99-01 Basis: Manual scram (trip) actions taken at the reactor control console are any set of actions by the Reactor Operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor. This condition indicates failure of the automatic protection system to scram the reactor. This condition Is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient. Thus, the plant safety has been compromised because design limits of the fuel may have been exceeded. An Alert is indicated because conditions may exist that lead to potential loss of fuel clad or RCS and because of the failure of the Reactor Protection System to automatically shut down the plant. If manual actions taken at the reactor control console fail to shut down the reactor, the event would escalate to a Site Area Emergency. PROCEDURE 5. 7 .1 REVISION 67 PAGE 220 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Automatic scram fails to shut down the reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor EAL: SS2.1 Site Area Emergency An automatic scram failed to shut down the reactor AND Manual actions taken at the reactor control console (NOTE 5) do not shut down the reactor as indicated by reactor power~ 3% NOTE 5 - Manual scram methods for EAL SA2.1 and EAL SS2.1 are the following:

Reactor Scram pushbuttons.
Reactor Mode switch in SHUTDOWN.
Manual or auto actuation of ARI.

Mode Applicability: 1 - Power Operation, 2 - Startup CNS Basis: This EAL addresses any automatic reactor scram signal followed by a manual scram 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. Methods of inserting a manual scram are limited to those that can be taken rapidly at the reactor control console (Panel 9-5) and include (Reference 1):

Both manual Reactor Scram pushbuttons.
Reactor Mode switch in SHUTDOWN.
Manual or auto actuation of ARI.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 221 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Auto actuation of ARI is included in the list of methods because the Operator, by procedure, always ensures actuation of ARI has occurred if the ARI actuation setpoints are exceeded. This means action to depress the ARI pushbuttons is taken if the automatic ARI actuation setpoints are exceeded but failed to actuate. If ARI properly actuates automatically, the ARI pushbuttons are not depressed. Reactor shutdown achieved by use of the alternate rod insertion methods listed in Emergency Procedure 5.8.3 do not constitute a successful manual scram (Reference 2). The APRM downscale trip setpoint (3%) is a minimum reading on the power range scale that indicates 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 the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine if reactor power is greater than 3% power (Reference 1, 3). Escalation of this event to a General Emergency would be under EAL SG2.1 or Emergency Director judgment. CNS Basis Reference(s):

1. General Operating Procedure 2.1.5, Reactor Scram.

2. Emergency Procedure 5.8.3, Alternate Rod Insertion Methods.

3. AMP-TBD00, Appendix B, Step RC/Q-4.

NEI 99-01 Basis: Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Manual scram actions taken at the reactor control console are any set of actions by the Reactor Operator(s) at which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 222 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Manual scram actions are not considered successful if action away from the reactor control console Is required to scram the reactor. This EAL ls still applicable even if actions taken away from the reactor control console are successful in shutting the reactor down because the design limits of the fuel may have been exceeded or because of the gross failure of the Reactor Protection System to shut down the plant. Escalation of this event to a General Emergency would be due to a prolonged condition leading to an extreme challenge to either core-cooling or heat removal. PROCEDURE 5.7.1 REVISION 67 PAGE 223 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Automatic scram and all manual actions fail to shut down the reactor and indication of an extreme challenge to the ability to cool the core exists EAL: SG2.1 General Emergency Automatic and all manual scrams were not successful AND Reactor power ~ 3% AND EITHER of the following exist or have occurred due to continued power generation: RPV level cannot be restored and maintained > -183 in. or cannot be determined OR Average torus water temperature and RPV pressure cannot be maintained within the Heat Capacity Temperature Limit (EOP/SAG Graph 7) Mode Applicability: 1 - Power Operation, 2 - Startup CNS Basis: This EAL addresses the following:

Any automatic reactor scram signal followed by failure of the automatic scram and all subsequent manual scrams 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 552.1); and
Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 224 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Reactor shutdown achieved by use of the alternate rod insertion methods listed in Procedure 5.8.3 are credited as a successful manual scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist (Reference 1). The APRM downscale trip setpoint is a minimum reading on the power range scale that indicates 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 the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine if reactor power is > 3% power (Reference 1, 3). 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 RCS barriers. Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above -183 inches (or cannot be determined). -183 inches is the Minimum Steam Cooling RPV Water Level (MSCRWL). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F (Reference 5). This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence. A threshold prescribing declaration when a parameter value cannot be maintained above a specified limit does not require immediate action simply because the current value is below the limit (-183 inches), but does not permit extended operation below the limit. Because the systems used to restore water level have different volume and discharge pressure capabilities, and pressure control methods also have varying rates that RPV pressure can be modified to support injection, the threshold must be considered reached as soon as it is apparent the limit cannot be attained within a reasonable amount of time. Determination of inability to restore and maintain RPV level is based on actions driven by EOPs to restore level. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 225 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] When RPV level cannot be determined, EOPs require entry to EOP-7B, RPV Flooding (Failure-to-Scram). RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The Instructions in EOP-7B specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures. The Heat Capacity Temperature Limit (HCTL) is the highest torus temperature from which Emergency RPV Depressurization will not raise torus pressure above the Primary Containment Pressure Limit (PCPL), while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCTL Is a function of RPV pressure and torus level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when EOP-3A, Primary Containment Control, Step SP/T-5, is reached (Reference 4). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. CNS Basis Reference(s):

1. General Operating Procedure 2.1.5, Reactor Scram.

2. Emergency Procedure 5.8.3, Alternate Rod Insertion Methods.

3. AMP-TBD00 Appendix B, Step RC/Q-4.

4. EOP-3A, Primary Containment Control.

5. NEDC 97-090]. .

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 226 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. The extreme challenge to the ability to cool the core Is intended to mean the reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level as described in the EOP bases. Considerations include inability to remove heat via the main condenser or via the suppression pool or torus (e.g., due to high pool water temperature). In the event either of these challenges exists at a time the reactor has not been brought below the power associated with the safety system design, a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier table declaration to permit maximum off-site Intervention time. PROCEDURE 5. 7 .1 REVISION 67 PAGE 227 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory:

3 - Inability to Reach Shutdown Conditions Initiating Condition: Inability to reach required shutdown within Technical Specification limits EAL:

SU3.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO action statement time Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: None CNS Basis Reference(s):

1. Technical Specifications.

NEI 99-01 Basis: Limiting Conditions of Operation (LCOs) require the plant to be brought to a required Shutdown Mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a 4 hour report under 10CFRS0. 72(b) non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCD-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other EALs. PROCEDURE 5. 7 .1 REVISION 67 PAGE 228 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 4 - Instrumentation/Communications Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room for 15 minutes or longer EAL: SU4.1 Unusual Event Unplanned loss of > ~ 75% of annunciators or indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C for~ 15 min. (NOTE 3) NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will _Ii kely _exceed_ the_ applicable_ ti me. _____________________________________ --------------------------------- _________ _ Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: The availability of computer-based monitoring capability (i.e., PMIS, SPDS) is not a factor at the Unusual Event emergency classification level. Safety system annunciation and indication considered in this EAL is found on Control Room Panels 9-3, 9-4, 9-5, and C. The other annunciators and indicators are important to plant operation but are not important to safety (Reference 1-14). CNS Basis Reference(s):

1. System Operating Procedure 2.2.64, Annunciator System.

2. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1.

3. Alarm Procedure 2.3_9-3-2, Panel 9 Annunciator 9-3-2.

4. Alarm Procedure 2.3_9-3-3, Panel 9 Annunciator 9-3-3.

5. Alarm Procedure 2.3_9-4-1, Panel 9 Annunciator 9-4-1.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 229 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

6. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2.

7. Alarm Procedure 2.3_9-4-3, Panel 9 Annunciator 9-4-3.

8. Alarm Procedure 2.3_9-5-1, Panel 9 Annunciator 9-5-1.

9. Alarm Procedure 2.3_9-5-2, Panel 9 Annunciator 9-5-2.

10. Alarm Procedure 2.3_C-1, Panel C - Annunciator C-1.

11. Alarm Procedure 2.3_C-2, Panel C - Annunciator C-2.

12. Alarm Procedure 2.3_C-3, Panel C - Annunciator C-3.

13. Alarm Procedure 2.3_C-4, Panel C - Annunciator C-4.

14. Abnormal Procedure 2.4ANN, Annunciator Failure.

NEI 99-01 Basis: This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Recognition of the availability of computer based indication equipment is considered. "PLANNED" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification is arbitrary, however, it is estimated that if 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 230 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific or several safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported per 10CFRS0. 72. If shutdown is not in compliance with the Technical Specification action, the UE is based on SU3.1, Inability to Reach Required Shutdown Within Technical Specification Limits. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. This UE will be escalated to an Alert based on a concurrent loss of compensatory indications or if a SIGNIFICANT TRANSIENT is in-progress during the loss of annunciation or indication. PROCEDURE 5. 7 .1 REVISION 67 PAGE 231 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 4 - Instrumentation Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room with EITHER (1) a significant transient in-progress, or (2) compensatory indicators unavailable EAL: SA4.1 Alert Unplanned loss of > ~ 75% of annunciators or indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C (NOTE 3) for~ 15 min. (NOTE 3) AND EITHER: Any significant transient is in-progress, Table 5-1 OR Compensatory indications are unavailable NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will _Ii ke ly _exceed_ the_ a pp Ii cab Ie _time. _______________________________________________________________________________ _ Table 5-1 Significan~ Transients Reactor scram Runback > 25% thermal power Electrical load rejection > 25% full electrical load ECCS injection Thermal power oscillations > 10% Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 232 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: PMIS and SPDS serve as redundant compensatory indicators which may be utilized in lieu of normal Control Room indicators. Safety system annunciation and indication considered in this EAL is found on Control Room Panels 9-3, 9-4, 9-5, and C. The other annunciators and indicators are important to plant operation but are not important to safety (Reference 1-14). Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as scrams, runbacks involving > 25% thermal power change, electrical load rejections of > 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater. CNS Basis Reference{s):

1. System Operating Procedure 2.2.64, Annunciator System.

2. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1.

3. Alarm Procedure 2.3_9-3-2, Panel 9 Annunciator 9-3-2.*

4. Alarm Procedure 2.3_9-3-3, Panel 9 Annunciator 9-3-3.

5. Alarm Procedure 2.3_9-4-1, Panel 9 Annunciator 9-4-1.

6. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2.

7. Alarm Procedure 2.3_9-4-3, Panel 9 Annunciator 9-4-3.

8. Alarm Procedure 2.3_9-5-1, Panel 9 Annunciator 9-5-1.

9. Alarm Procedure 2.3_9-5-2, Panel 9 Annunciator 9-5-2.

10. Alarm Procedure 2.3_C-1, Panel C - Annunciator C-1.

11. Alarm Procedure 2.3_C-2, Panel C - Annunciator C-2.

12. Alarm Procedure 2.3_C-3, Panel C - Annunciator C-3.

13. Alarm Procedure 2.3_C-4, Panel C - Annunciator C-4.

14. Abnormal Procedure 2.4ANN, Annunciator Failure.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 233 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES

                      ,  [INFORMATION USE]

NEI 99-01 Basis: This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a SIGNIFICANT TRANSIENT. "PLANNED" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification is arbitrary, however, it is estimated that if 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide Increased monitoring of system operation. It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific or several safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported per 10CFRS0. 72. If shutdown Is not in compliance with the Technical Specification action, the UE is based on EAL SU3.1, Inability to Reach Required Shutdown Within Technical Specification Limits. "Compensatory indications" in this context includes computer based information such as PMIS/SPDS. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. This Alert will be escalated to a Site Area Emergency if Operating Crew cannot monitor the transient in-progress due to a concurrent loss of compensatory indications with a SIGNIFICANT TRANSIENT in-progress during the loss of annunciation or indication. PROCEDURE 5.7.1 REVISION 67 PAGE 234 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 4 - Instrumentation Initiating Condition: Inability to monitor a significant transient in-progress EAL: SS4.1 Site Area Emergency Loss of > ~ 75% of the annunciators or indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C for~ 15 min. (NOTE 3) AND Any significant transient is in-progress, Table S-1 AND Compensatory indications are unavailable NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will _Ii kely _exceed_ the_ a pp Ii cab Ie _time. _______________________________________________________________________________ _ Table S-1 Significant Transients Reactor scram Runback > 25% thermal power Electrical load rejection > 25% full electrical load ECCS injection Thermal power oscillations > 10% Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 235 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: The availability of computer-based monitoring capability (i.e., PMIS, SPDS) is a factor at the Site Area Emergency classification level because they are compensatory non-alarming indication. Safety system annunciation and indication considered in this EAL is found on Control Room Panels 9-3, 9-4, 9-5, and C. The other annunciators and indicators are important to plant operation but are not important to safety (Reference 1-14). Significant transients are listed in Table 5-1 and include response to automatic or manually initiated functions such as trips, runbacks involving > 25% thermal power change, electrical load rejections of > 25% full electrical load, ECCS injections, or thermal power oscillations of> 10%. Due to the limited number of safety systems in operation during Cold Shutdown, Refueling and Defueled Modes, this EAL is not applicable during these modes of operation. CNS Basis Reference(s):

1. System Operating Procedure 2.2.64, Annunciator System.

2. Alarm Procedure 2.3_9-3-1, Panel 9 Annunciator 9-3-1.

3. Alarm Procedure 2.3_9-3-2, Panel 9 Annunciator 9-3-2.

4. Alarm Procedure 2.3_9-3-3, Panel 9 Annunciator 9-3-3.

5. Alarm Procedure 2.3_9-4-1, Panel 9 Annunciator 9-4-1.

6. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2.

7. Alarm Procedure 2.3_9-4-3, Panel 9 Annunciator 9-4-3.

'8. Alarm Procedure 2.3_9-5-1, Panel 9 Annunciator 9-5-1.

9. Alarm Procedure 2.3_9-5-2, Panel 9 Annunciator 9-5-2.

10. Alarm Procedure 2.3_C-1, Panel C - Annunciator C-1.

11. Alarm Procedure 2.3_C-2, Panel C - Annunciator C-2.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 236 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

12. Alarm Procedure 2.3_C-3, Panel C - Annunciator C-3.

13. Alarm Procedure 2.3_C-4, Panel C - Annunciator C-4.

14. Abnormal Procedure 2.4ANN, Annunciator Failure.

NEI 99-01 Basis: This EAL is intended to recognize the threat to plant safety associated with the complete loss of capability of the Control Room Staff to monitor plant response to a SIGNIFICANT TRANSIENT. "PLANNEO" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor. Quantification is arbitrary; however, it is estimated that if~ 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation. It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific or several safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported per 10CFRSO. 72. If shutdown is not in compliance with the Technical Specification action, the UE is based on SU3.1, Inability to Reach Required Shutdown Within Technical Specification Limits. A Site Area Emergency is considered to exist if Control Room Staff cannot monitor safety functions needed for protection of the public while a significant transient is in-progress. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 237 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Site specific indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications, and dedicated annunciation capability. "Compensatory indications" in this context includes computer based information such as PMIS/SPDS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. PROCEDURE 5. 7.1 REVISION 67 PAGE 238 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 5 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL: SU5.1 Unusual Event SJAE monitor > 1.58E+3 mR/hr Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: Steam Jet Air Ejectors (SJAEs) remove all non-condensable gases from the condensers including air in-leakage and disassociated products originating in the reactor and exhausts them to the off-gas holdup volume. A rise in off-gas activity could therefore indicate damage to the fuel cladding, a potential degradation in the level of safety of the plant, and a potential precursor of more serious problems. The Technical Specification allowable limit is ~ 1 Ci/sec. The SJAE monitor Hi-Hi radiation setpoint is set at 50% of the instantaneous release limit and represents ~ 0.1 % fuel cladding damage. The SJAE monitor Hi-HI radiation setpoint has been selected because it is operationally significant and is readily recognizable by the Control Room Operating Staff (Reference 2-6). The Off-Gas System isolates after a 15 minute time delay (Reference 2, 3). In the Hot Modes, a steam source is available from which non-condensables can be separated for processing by the Off-Gas System. The Cold Shutdown, Refueling, and Defueled Modes do not afford a transfer mechanism from which the off-gas radiation monitors can draw a valid sample. The radiation monitors lose a valid sample source when the air ejectors are not in service (Reference 2, 4, 5). (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 239 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. System Operating Procedure 2.2.55, Main Condenser Gas Removal System.

2. Alarm Procedure 2.3_9-4-1, Panel 9 Annunciator 9-4-1, C-4, OFFGAS TIMER INffiATED.

3. Alarm Procedure 2.3_9-4-1, Panel 9 Annunciator 9-4-1, C-5, OFFGAS HIGH RAD.

4. Abnormal Procedure 2.4OG, Off-Gas Abnormal.

5. Emergency Procedure 5.2FUEL, Fuel Failure.

6. NEDC 02-004, Estimation of the Steam Jet Air Ejector Radiation Monitor, RMP-RM-150A(B), Readings Following a 1% Fuel Clad Release (Degraded Core) in the Reactor Coolant System.

7. Technical Specification LCO 3.7.5, Air Ejector Off-Gas.

NEI 99-01 Basis: This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the pla'nt. Escalation of this EAL to the Alert level is per the Fission Product Barriers. This threshold addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity. PROCEDURE 5.7.1 REVISION 67 PAGE 240 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 5 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL: SUS.2 Unusual Event Coolant activity~ 4.0 µCi/gm dose equivalent 1-131 Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS 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. This EAL addresses reactor coolant samples exceeding Technical Specification LCO 3.4.6, which is applicable in Hot Operating Modes (Reference 1). CNS Basis Reference(s):

1. Technical Specification LCO 3.4.6.

NEI 99-01 Basis: This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. Escalation of this EAL to the Alert level is per the Fission Product Barriers. This threshold addresses coolant samples exceeding coolant Technical Specifications for transient iodine spiking limits. PROCEDURE 5. 7 .1 REVISION 67 PAGE 241 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 6 - RCS Leakage Initiating Condition: RCS leakage EAL: SU6. l Unusual Event Unidentified or pressure boundary leakage > 10 gpm OR Identified leakage > 30 gpm (NOTE 6) NOTE 6 - See Table F-1, Fission Product Barrier Matrix, for possible escalation a ~ove _the_ Un usu a I_ Event_ due _to_ RCS _Leakage.----------------------------------------------------------- Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: Leakage is monitored by utilizing the following techniques (Reference 1):

Sensing excess flow in piping systems.
Sensing pressure and temperature changes in the Primary Containment.
Monitoring for high flow and temperature through selected drains.
Sampling airborne particulate and gaseous radioactivity.
Drywell floor and equipment drain sump leak rate alarm system.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 242 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] The 10 gpm value for the unidentified drywell leakage was selected because it is observable with normal Control Room measurement of sump pump out rates (e.g., Drywell Sump Pump Flow RW-FR-528, red/blue pen, etc.). Drywell equipment Sump G and drywell floor drain Sump F each have a FILL UP RATE HIGH annunciator on Panel 9-4-2. If either sump fills from the low-level switch reset point to the high-level pump start point before a preset timer has timed out, the annunciator will alarm indicating the sumps are filling at an excessive rate. Sumps F and G will overflow to each other through a trench system. Drywell equipment and floor drain sump pump isolation valves isolate on RPV low water level (~ 3 inches) or high drywell pressure (~ 1.84 psig). A SRV that opens but cannot be closed from the Control Room meets this criterion and the UE should be declared. CNS Basis Reference(s):

1. System Operating Procedure 2.2.27, Equipment, Floor, and Chemical Drain System.

2. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2, B-1, DRYWELL EQUIP SUMP G HIGH FILL-UP RATE.

3. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2, B-2, DRYWELL FLOOR DRN SUMP F HI FILL-UP RATE.

4. Surveillance Procedure 6.LOG.601, Daily Surveillance Log - Modes 1, 2, and 3.

5. Technical Specification LCO 3.4.4, RCS Operational Leakage.

6. Technical Specification LCO 3.4.5, RCS Leakage Detection Instrumentation.

7. USAR Section X-14.0, Equipment and Floor Drainage Systems.

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 243 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with normal Control Room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). Relief valve normal operation should be excluded from this EAL. However, a relief valve that operates and fails to close per design should be considered applicable to this EAL if relief valve cannot be isolated. The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. In either case, escalation of this EAL to the Alert level is per the Fission Product Barrier Degradation EALs. PROCEDURE 5. 7.1 REVISION 67 PAGE 244 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 7 - Loss of DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS7 .1 Site Area Emergency

  < 105 VDC bus voltage indications on all vital 125 VDC buses (lA and lB) for
 ;: : 15 min. (NOTE 3)

NOTE 3 - The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined the condition will _Ii ke Iy _exceed_ the_ a pp Ii cab Ie _time. _______________________________________________________________________________ _ Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: 105 VDC is the minimum design bus voltage (Reference 4). The 125 VDC System supplies DC power to conventional station emergency equipment and selected Safeguard System loads. 125 VDC Distribution Panels supply control and instrument power for annunciators control logic power, and protective relaying. Figure S-2 illustrates the 125 VDC Power System (Reference 3). If 125 VDC Distribution Panel A is lost, the following major equipment is affected: RRMG A speed and breaker control, 4160V Bus lA, lE, and lF breaker control and undervoltage logics, 480V Bus lA and lF breaker control, the right light in all Control Room annunciators, annunciator panels for Water Treatment, RHR A Gland Water, Auxiliary Steam Boiler C, DG-1 starting and breaker control logics, CS A, RCIC, and RHR A control logics, TIP valve control monitors, main generator voltage regulation, RFPT A trip logic, and ARI solenoid valve power. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 245 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] If 125 VDC Distribution Panel B is lost, the following major equipment is affected: RRMG B speed and breaker control, 4160V Bus 1B and 1G breaker control and undervoltage logics, 480V Bus 1B and lG breaker control, the left light in all Control Room annunciators, annunciator panels for ALRW, RHR B Gland Water, Auxiliary Steam Boiler D, DG-2 starting and breaker control logics, CS B, HPCI, and RHR B control logics, main generator trip logic, main generator and transformer protective relaying, bypass valves fail to control pressure after turbine trip and RFPT B trip logic. Battery chargers receive their power from 460V critical motor control centers. Each 125 VDC bus receives power from either a 125 VDC battery or a 125 VDC battery charger. The battery chargers receive their power from 460V critical motor control centers. The 250 VDC System supplies DC power to conventional station emergency equipment and selected Safeguard System loads. Although 250 VDC Buses 1A and 1B provide vital DC emergency power, 250 VDC Safety System loads (such as motor operated valves) also require 125 VDC control power. Loss of 125 VDC buses alone, therefore, would render most Safeguard System loads inoperable (Reference 4, 5, 6). This EAL is the hot condition equivalent of the cold condition loss of DC power EAL CU6.1. CNS Basis Reference(s):

1. Emergency Procedure 5.3DC125, Loss of 125 VDC.

2. Surveillance Procedure 6.EE.607, 125V Station Battery Modified Performance Discharge Test.

3. BR 3058 DC One Line Diagram.

4. Technical Specifications B 3.8.4.

5. USAR Section Vlll-6.2.

6. USAR Section VIII-6.3.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 246 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Figure 5-2: 125 VDC Power System MCC-L MCC-TX (20) ---~-- TX I flexUSVOC Disconnect I I I I I I I I I I I GNO I MCC,-TX DET

LIGHTS:

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                                -    ---1               1---        --------------1 A I             I  A                                                          I I             I                                                             I I             I        -.:...._....:.,;.:.:::.:.;...:..=-,--1~---,-...,::s I I             I                                                             I I                                      I             I                                                             I I                                      I             I                                                             I N        E RCIC E

RxBLDG N e DIST. PANELA 126VDCSR"B" E N 6 !l DIST. PANEL B E HPCI PANEL HPCI M0-16 (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 24 7 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the Reactor System. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation. PROCEDURE 5.7.1 REVISION 67 PAGE 248 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: S - System Malfunction Subcategory: 8 - Communications Initiating Condition: Loss of all on-site or off-site communications capabilities EAL: SUS. l Unusual Event Loss of all Table S-2 on-site (internal) communications capability affecting the ability to perform routine operations OR Loss of all Table S-2 off-site (external) communications methods affecting the ability to perform off-site notifications Table 5-2 Communications Systems System Onsite Offsite (internal) (external) Station Intercom System "Gaitronics" X Site UHF Radio Consoles X Radio Paging System X Alternate Intercom X CNS On-Site Cell Phone System X X Telephone system (PBX) X X Federal Telecqmmunications System (FTS 2001) X Local Telephones (C.O. Lines) X CNS State Notification Telephones X Satellite Telephones X Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 249 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis: NOTE- EPIP 5.7COMMUN has more detail on each of the communications systems covered_ by_ this_ EAL.---------------------------------------------------------------------------------------------------- On-site/off-site communications include one or more of the systems listed in Table S-2 (Reference 1).

Station Intercom System "Gaitronics": Permits communication between the different parts of the plant and it also incorporates a public address system for plant wide announcements.
Site UHF Radio Consoles: The site UHF radio system uses four repeaters; Base 1 and Base 2 are used by Operations, Base 3 and Base 4 are used by Security. These repeaters operate on different frequencies. All remote control, portable, and mobile units are capable of selecting either repeater.
Radio Paging System: CNS leases pagers and radio paging services from a telecommunications company. Pagers are issued to various Management and Emergency Response personnel at CNS and other NPPD locations. Pagers can be activated from any touch-tone phone, on-site or off-site.
Alternate Intercom: Provides an alternate in-plant communications network utilizing a secondary telephone system. This system is located in the ERP shack and has battery back-up.
CNS On-Site Cell Phone System.
Telephone System (PBX): Provides voice communication between virtually all buildings, offices, and operation facilities within the station. The telephone system also provides communications between the plant and off-site facilities via the telephone switchboard network. The system allows Operating Crews to alert plant personnel in emergencies. The telephone company provides the normal and leased line services.
Federal Telecommunications System (FTS 2001): The Health Physics Network (HPN) and Emergency Notification System (ENS) provide communications between NRC and CNS during an emergency.
Local Telephones (C.O. Lines).

(continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 250 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE]

CNS State Notification Telephones: The CNS State Notification Telephone System is the primary means for the plant to make emergency notifications to state and local authorities. This system provides direct communication with the Nebraska State Patrol, the Missouri State Patrol, the Atchison County Sheriff's Department, and the Nemaha and Richardson County Sheriff's Departments.
Satellite Telephones.

This EAL is the hot condition equivalent of the cold condition EAL CU4.1. CNS Basis Reference(s):

1. EPIP 5. 7COMMUN, Communications, Emergency Response Facility Communication Equipment attachment.

NEI 99-01 Basis: This EAL addresses loss of communications capability that either prevents the plant operations staff ability to perform routine tasks necessary for plant operations or inhibits the ability to communicate problems externally to off-site authorities from the Control Room. The loss of off-site communications ability encompasses the loss of all means of communications with off-site authorities and is expected to be significantly more comprehensive than the condition addressed by 10CFR50. 72. The availability of one method of ordinary off-site communications is sufficient to inform federal, state, and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to off-site locations, etc.) are being utilized to make communications possible. PROCEDURE 5. 7 .1 REVISION 67 PAGE 251 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: E - ISFSI Subcategory: None Initiating Condition: Damage to a loaded cask confinement boundary EAL: EU 1.1 Unusual Event Damage to a loaded cask confinement boundary Mode Applicability: N/A CNS Basis: Minor surface damage that does not affect storage cask boundary is excluded from the scope of this EAL. CNS Basis Reference(s):

1. Certifc!cate of Compliance Number 1004, April 17, 2007.

N EI 99-01 Basis: An Unusual Event in this EAL is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on a loaded fuel storage cask confinement boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage. PROCEDURE 5. 7.1 REVISION 67 PAGE 252 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of Primary Containment EAL: FUl.1 Unusual Event Any loss or any potential loss of Primary Containment (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: Fuel Clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 3) lists the fission product barrier thresholds, bases, and references. Fuel Clad and RCS barriers are weighted more heavily than the Primary Containment barrier. Unlike the Fuel Clad and RCS barriers, the loss of either of which results in an Alert (EAL FAl.1), loss of the Primary Containment barrier in and of itself does not result in the relocation of radioactive materials or the potential for degradation of core cooling capability. However, loss or potential loss of the Primary Containment barrier in combination with the loss or potential loss of either the Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FSl.1. CNS Basis Reference(s): None NEI 99-01 Basis: None PROCEDURE 5. 7.1 REVISION 67 PAGE 253 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: Fuel Clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 3) lists the fission product barrier thresholds, bases, and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Primary Containment barrier. Unlike the Primary Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Primary Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1. CNS Basis Reference(s): None NEI 99-01 Basis: None PROCEDURE 5.7.1 REVISION 67 PAGE 254 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FSl.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: Fuel Clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 3) lists the fission product barrier thresholds, bases, and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

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

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds In addition to off-site dose assessments would require continual assessments of radioactive inventory and Primary Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 255 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s): None NEI 99-01 Basis: None PROCEDURE 5. 7 .1 \ REVISION 67 PAGE 256 OF 342

ATTACHMENT 2 EMERGENCY ACTION LEVEL TECHNICAL BASES [INFORMATION USE] Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FGl.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown CNS Basis: Fuel Clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 3) lists the fission product barrier thresholds, bases, and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS, and Primary Containment barriers.
Loss of Fuel Clad and RCS barriers with potential loss of Primary Containment barrier.
Loss of RCS and Primary Containment barriers with potential loss of Fuel Clad barrier.
Loss of Fuel Clad and Primary Containment barriers with potential loss of RCS barrier.

CNS Basis Reference(s): None N EI 99-01 Basis: None PROCEDURE 5.7.1 REVISION 67 PAGE 257 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] ATTAO-IMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TEO-INICAL BASES [INFORMATION USE] 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 Primary 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. RPV Level. B. PC Pressure/Temperature. C. Isolation. D. ERD. E. Rad. F. 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 so that they can be easily identified. 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. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 258 OF 342 J I

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] 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 EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Primary Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given In EALs FGl.1, FSl.1, FAl.1, and FU 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 RCS barrier, and finally the Primary Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B, then C, then D, then E, and then F. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 259 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Tiwle.f=-1' llssJgn Pr9d~) B.arrlerMat~x~ I Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Loss Potential Loas Loss Pote ntlal Los11 Lo8II Potential Lose Alti'\'UI ..... 1 SAG l~1.1req.ireddiMIIO_,.,.

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ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

1. SAG 1 entry is required due to any of the following:

Non-Failure-to-Scram:

o RPV water level cannot be restored and maintained > -183 in., or o RPV water level cannot be restored and maintained ~ -209 in. and no core spray subsystem flow can be restored and maintained

           ~ 4,750 gpm

Failure-to-Scram:

o RPV water level cannot be restored and maintained > -183 in. and core steam flow cannot be restored and maintained > 800,000 lbm/hr

Core damage is occurring due to loss of core cooling.

CNS Basis: EOP-lA, EOP-2B; EOP-7A, and EOP-7B specify entry into SAG 1 when it is determined that core damage is occurring due to loss of core cooling (Reference 1). SAG entry signifies the need to implement severe accident mitigation actions. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAG 1 entry is required during Non-Failure-to-Scram events when any of the following conditions exist (Reference 1):

RPV water level cannot be restored and maintained above -183 inches (MSCRWL, EOP-lA) (Reference 2, 6).
RPV water level cannot be restored and maintained at or above -209 inches (elevation of the jet pump suction) and no core spray subsystem flow can be restored and maintained equal to or~ 4,750 gpm (design core spray flow, EOP-lA) (Reference 2, 7).

( continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 261 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] SAG 1 entry ~s required during Failure-to-Scram events when RPV water level cannot be restored and maintained > -183 inches and core steam flow cannot be restored and maintained > 800,000 lbm/hr (Reference 3, 6). The specified steam flow is the Minimum Core Steam Flow (MCSF). The MCSF is the lowest core steam flow sufficient to preclude any clad temperature from exceeding 1500°F even if reactor core is not completely covered (Reference 6). The MCSF is only applicable in failure-to-scram events because reactor power must be well above the decay heat generation rate for steam production to equal the MCSF. Whether or not RPV water level can be determined, SAG 1 entry is also required when core damage is occurring due to loss of core cooling (EOP-lA, EOP-2B, EOP-7A, or EOP-7B) (Reference 2, 3, 4, 5). If RPV water level cannot be determined, the absence of core damage indications may be the only means of determining if adequate core cooling is being maintained. If RPV water level can be determined, restoration of RPV water level to above -183 inches and restoration of core spray cooling requirements may not occur in a timely manner. If indications of core damage occur while RPV injection is being restored, entry to SAG 1 is appropriate even if required water levels and spray cooling flow are eventually achieved. Fuel Clad Losses 2 and 6 (drywell radiation), 3 (primary containment activity), 4 (MSL Hi-Hi Rad), and 5 (SJAE Monitor dose rate) are thresholds that would be indicative of core damage occurring due to loss of core cooling. The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad. This threshold is also a Potential Loss of the Primary Containment barrier (PC P-Loss 25). Since the EOP requirement for SAG 1 entry is reached after core uncovery has occurred, a Loss of the RCS barrier exists (RCS Loss 10). SAG entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. The EALs are aligned to the updated guidance of EPG/SAG Revision 3 such that the plant Operator can readily implement both when needed (Reference 8). ( continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 262 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. AMP-TBD00 PSTG/SATG Technical Bases, RC/L, Contingency #1, #4, #5.

2. EOP-lA, RPV Control.

3. EOP-7A, RPV Level (Failure-to-Scram).

4. EOP-2B, RPV Flooding.

5. EOP-7B, RPV Flooding (Failure-to-Scram).

6. NEDC 97-090].

7. NEDC 97-089.

8. EPFAQ Number 2015-004.

NEI 99-01 Basis: The "Loss" threshold value corresponds to the level used in EOPs to indicate challenge of core cooling. This is the minimum value to assure core cooling without further degradation of the clad. PROCEDURE 5. 7 .1 REVISION 67 PAGE 263 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

8. RPV level cannot be restored and maintained > -158 in. or cannot be determined CNS Basis:

The KEY consideration of this criterion is adequate core cooling. RPV conditions that result In Inadequate core cooling shall result in determinations that the fuel clad barrier has been potentially lost and the RCS barrier has been lost. A RPV level instrument reading of -158 inches indicates RPV level is at the top of active fuel (TAF) (Reference 4). When RPV level is at or above TAF, the core is completely submerged. Core submergence Is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). To reach this level, RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If after RPV pressure reduction (either manually, automatically, or by failure of the RCS barrier), RPV level cannot be restored and maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops below TAF, the level ls indicative of a challenge to core cooling and the Fuel Clad barrier. EOPs allow the Operator a wide choice of RPV injection sources to consider when evaluating if RPV water level can be restored and maintained to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, an important consideration Is the status of adequate core cooling during the pressure reduction process. A rapid depressurization which allows refilling the RPV with a high volume system will normally maintain adequate cooling with steam flow. Slow depressurizations, however, may challenge adequate core cooling due to both low water levels and low steam flows. (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 264 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] A threshold prescribing declaration when a parameter value cannot be maintained above a specified limit does not require immediate action simply because the current value is below the limit (top of active fuel), but does not permit extended operation below the limit. Because the systems used to restore water level have different volume and discharge pressure capabilities, and pressure control methods also have varying rates that RPV pressure can be modified to support injection, the threshold must be considered reached as soon as it is apparent the top of active fuel cannot be attained within a reasonable amount of time. Determination of inability to restore and maintain RPV level is based on actions driven by EOPs to restore level. In* high-power ATWS/failure to scram events, EOPs may direct the Operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power Is the higher priority. For such events, ICs SA2, 552, and SG2 will dictate an emergency classification. The fission product barrier criteria should continue to be evaluated independently to identify barrier conditions that would require escalation of the classification. When RPV level cannot be determined, EOPs require entry to EOP-2B, RPV Flooding, or EOP-7B, RPV Flooding (Failure-to-Scram). RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-2B/7B specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists (Reference 1, 3). CNS Basis Reference(s):

1. EOP-2B, RPV Flooding.

2. EOP-7A, RPV Level (Failure-to-Scram).

3. EOP-7B, RPV Flooding (Failure-to-Scram).

4. NEDC 97-089.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 265 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This threshold is the same as the RCS barrier "Loss" Threshold A.10 and corresponds to the water level at the top of the active fuel. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. PROCEDURE 5.7.1 REVISION 67 PAGE 266 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: E. Rad Degradation Threat: Loss Threshold: I2. Drywell radiation monitor (RMA-RM-40A/B) > 2.50E+03 Rem/hr CNS Basis: EPIP 5.7.17.1, Dose Assessment (Manual), Core Damage Estimation attachment provides a method of calculating percent fuel clad damage and fuel melt based on drywell radiation. Under LOCA conditions, a reading of 2.44E+6 Rem/hr corresponds to 100% core melt on drywell radiation Monitors RMA-RM-40A/B. A value of 2.44E+3 Rem/hr (rounded to 2.50E+03 Rem/hr) yields 1% fuel clad damage using this method. In order to reach this Fuel Clad barrier Potential Loss threshold, a loss of the RCS barrier has already occurred (see RCS Loss 14). This threshold, therefore, represents at least a Site Area Emergency classification. CNS Basis Reference{s):

1. EPIP 5.7.17.1, Dose Assessment (Manual).

NEI 99-01 Basis: 2.50E+03 Rem/hr is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration

of~ 300 µCi/gm dose equivalent I-131 ( 1% clad damage) into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within Technical Specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss Threshold 14. Thus, this threshold indicates a loss of both Fuel Clad barrier and RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. PROCEDURE 5. 7 .1 REVISION 67 PAGE 267 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: E. Rad Degradation Threat: Loss Threshold: I3. Primary coolant activity > 300 µCi/gm dose equivalent I-131© 4 CNS Basis: None CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. NEI 99-01, Revision 5.

3. NEDC 02-20, Estimation of Reactor Coolant System Dose Equivalent I-131 Concentration Following a 1% Fuel Clad Failure (Degraded Core) Under Non-LOCA Conditions.

NEI 99-01 Basis: Coolant activity of 300 µCi/gm dose equivalent I-131 is well above that expected for iodine spikes and corresponds to about 1% fuel clad damage. This amount of radioactivity Indicates significant clad damage and thus, the Fuel Clad Barrier is considered lost. PROCEDURE 5.7.1 REVISION 67 PAGE 268 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: - Fuel Clad Category: E. Rad Degradation Threat: Loss Threshold:

4. Main Steam Line Radiation Monitor Readings ~ Hi-Hi Alarm Setpoint CNS Basis:

The Hi-Hi alarm setpoint for the Main Steam Line Radiation Monitors is based on a Control Rod Drop Accident. This accident is most severe when initiated at< 10% rated thermal power. The setpoint is a multiple of the normal full power background reading on these monitors that was observed during the previous operating cycle. Clad damage resulting in the Hi-Hi Main Steam Line Radiation Monitor Alarms due to a Control Rod Drop Accident could exceed 2% of the core's fuel rods. USAR XIV-6.2 discusses the Control Rod Drop Accident. Specific Emergency Director evaluation is required to determine if the requisite plant conditions exist that would make this criteria valid.© 3 CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. USAR XIV-6.2, Control Rod Drop Accident.

NEI 99-01 Basis: None PROCEDURE 5.7.1 REVISION 67 PAGE 269 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: E. Rad Degradation Threat: Loss Threshold: I5. ~ l.5E4 mrem/hr on SJAE Monitor CNS Basis: NEDC 02-004 assumes that cladding damage has resulted in 300 µCi/gm of I-131 In the reactor coolant, that non-LOCA conditions exist, and that normal sample flow through the monitor is maintained. Under these conditions, the SJAE monitors are expected to reach 15,000 mrem/hr. If sample flow is lost or the monitors are isolated and they reach this point, they are assumed to have failed. Specific Emergency Director evaluation is required to determine if the requisite plant conditions exist that would make this criteria valid.© 3 CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. NEDC 02-004, Estimation of the Steam Jet Air Ejector Radiation Monitor RMP-RM-150A(B), Readings Following a 1% Fuel Clad Release (Degraded Core) In the Reactor Coolant System.

NEI 99-01 Basis: None PROCEDURE 5.7.1 REVISION 67 PAGE 270 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: E. Rad Degradation Threat: Loss Threshold:

16. Non-LOCA with DW Rad Monitor reading > 115 REM/hr CNS Basis:

NEDC 02-009 assumes that cladding damage has resulted in 300 µCi/gm of I-131 in the reactor coolant and that non-LOCA conditions exist. The calculated value is based strictly on shine from the RR System piping, and as a result, is only valid if primary containment parameters do not indicate increased RCS leakage. With increased RCS leakage, any contaminants in the coolant could be in closer proximity to the detectors (as an aerosol or vapor) causing the monitors to read higher. If elevated RCS leakage is suspected, a drywell radiation monitor reading of 2500 R/hr, per Criteria 2, is the appropriate threshold for this parameter to determine Fuel Clad failure. Specific Emergency Director evaluation is required to determine if the requisite plant conditions exist that would make this criteria valid.© 3 CNS Basis Reference(s):

1. EPIP 5.7.17, CNS-DOSE Assessment, and/or EPIP 5.7.17.1, Dose Assessment (Manual).

2. NEDC 02-009, Estimation of Primary Containment High Range Monitor RMA RM 40A(B), Readings Following 1% Clad Failure in the RCS Under Non-LOCA Conditions.

NEI 99-01 Basis: None PROCEDURE 5.7.1 REVISION 67 PAGE 271 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: F. Judgment Degradation Threat: Loss Threshold:

7. Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barri~r CNS Basis:

The Emergency Director judgment threshord 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 2 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 off-site monitoring results.
Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the

  , Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CNS Basis Reference(s): None NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. In addition, the inability to monitor the barrier should also be considered as a factor in Emergency Director judgment that the barrier may be considered lost. PROCEDURE 5.7.1 REVISION 67 PAGE 272 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Fuel Clad Category: F. Judgment Degradation Threat: Potential Loss Threshold:

9. Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier CNS Basis:

The Emergency Director 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 2 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 off-site monitoring results.
Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CNS Basis Reference(s): None NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. In addition, the inability to monitor the barrier should also be considered as a factor in Emergency Director judgment that the barrier may be considered potentially lost. PROCEDURE 5. 7 .1 REVISION 67 PAGE 273 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Loss Threshold:

10. RPV level cannot be restored and maintained > -158 in. or cannot be determined CNS Basis:

The KEY consideration of this criterion is the ability to maintain adequate core cooling. RPV conditions that result in inadequate core cooling shall result In determinations that the fuel clad barrier has been potentially lost and the RCS barrier has been lost. A RPV level instrument reading of -158 inches indicates RPV level is at the top of active fuel (TAF) (Reference 4). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV Inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If after RPV pressure reduction (either manually, automatically, or by failure of the RCS barrier), RPV level cannot be restored and maintained above TAF, ECCS, and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier. EOPs allow the Operator a wide choice of RPV injection sources to consider when evaluating if RPV water level can be restored and maintained, to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, an important consideration is the status of adequate core cooling during the pressure reduction process. A rapid depressurization which allows refilling the RPV with a high volume system will normally maintain adequate cooling with steam flow. Slow depressurizations, however, may challenge adequate core cooling due to both low water levels and low steam flows. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 274 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] A threshold prescribing declaration when a parameters value cannot be maintained above a specified limit does not require immediate action simply because the current value is below the limit (top of active fuel), but does not permit extended operation below the limit. Because the systems used to restore water level have different volume and discharge pressure capabilities, and pressure control methods also have varying rates that RPV pressure can be modified to support injection, the threshold must be considered reached as soon as it is apparent the top of active fuel cannot be attained within a reasonable amount of time. Determination of inability to restore and maintain RPV level is based on actions driven by EOPs to restore level. In high-power ATWS/failure to scram events, EOPs may direct the Operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA2, SS2, and SG2 will dictate an emergency classification. The fission product barrier criteria should continue to be evaluated independently to identify barrier conditions that would require escalation of the classification. When RPV level cannot be determined, EOPs require entry to EOP-2B, RPV Flooding, or EOP-7B, RPV Flooding (Failure-to-Scram). The instructions in EOP-2B/7B specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss 10) (Reference 1, 3). CNS Basis Reference(s):

1. EOP-2B, RPV Flooding.

2. EOP-7A, RPV Level (Failure-to-Scram).

3. EOP-7B, RPV Flooding (Failure-to-Scram).

4. NEDC 97-089.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 275 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES JINFORMATION USE] N EI 99-01 Basis: The Loss threshold for RPV water level corresponds to the level that is used in EOPs to indicate challenge of core cooling. This threshold is the same as Fuel Clad Barrier Potential Loss Threshold #8 and corresponds to a challenge to core cooling. Thus, this threshold indicates a Loss of RCS barrier and Potential Loss of Fuel Clad barrier that appropriately escalates the emergency classification level to a Site Area Emergency. There is no Potential Loss threshold associated with this item. PROCEDURE 5.7.1 REVISION 67 PAGE 276 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: B. PC Pressure/Temperature Degradation Threat: Loss Threshold: I11. PC pressure > 1.84 psig due to 'Res leakage CNS Basis: The Primary Containment (PC) high pressure scram setpoint is an entry condition to EOP-1A, RPV Control, and EOP-3A, Primary Containment Control (Reference 1, 2). Normal Primary Containment pressure control functions (e.g., operation of drywell cooling, SBGT, etc.) are specified in EOP-3A in advance of less desirable but more effective functions (e.g., operation of drywell or torus sprays, etc.). In the CNS design basis, Primary Containment pressures above the high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control Primary Containment vent/purge (Reference 3). The threshold phrase " ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect Primary Containment pressure. PC pressure > 1.84 psig with corollary indications (drywell temperature, humidity, etc.) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure > 1.84 psig should not be considered a RCS barrier Loss. CNS Basis Reference(s):

1. EOP-1A, RPV Control.

2. EOP-3A, Primary Containment Control.

3. USAR Section XIV-6.3.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 277 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The threshold pressure value is the Primary Containment high pressure scram setpoint and is indicative of a LOCA event that requires ECCS response. There is no Potential Loss threshold associated with this item. PROCEDURE 5.7.1 REVISION 67 PAGE 278 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Loss Threshold:

12. Release pathway exists outside Primary Containment resulting from isolation failure in any of the following (excluding normal process system flowpaths from an unisolable system):

Main steam line.
HPCI steam line.
RCIC steam line.
RWCU.
Feedwater.

CNS Basis: The conditions of this threshold Include required containment isolation failures allowing a flow path to the environment. A release pathway outside Primary Containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if emergency response requires the normal process flow of a system outside Primary Containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss 21) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater Systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 279 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. System Operating Procedure 2.2.33, High Pressure Coolant Injection System.

2. System Operating Procedure 2.2.56, Main Steam System.

3. System Operating Procedure 2.2.66, Reactor Water Cleanup.

4. System Operating Procedure 2.2.67, Reactor Core Isolation Cooling System.

5. BR 2041, Reactor Building Main Steam System.

6. BR 2042.

7. BR 2043.

8. BR 2044, HPCI System.

NEI 99-01 Basis: An unisolable RCS break outside Primary Containment is a breach of the RCS barrier. Thus, this threshold is included for consistency with the Alert emergency classification level. Large high-energy line breaks such as HPCI, Feedwater, RWCU, or RCIC that are unisolable represent a significant loss of the RCS barrier and should be considered as MSL breaks for purposes of classification. PROCEDURE 5. 7 .1 REVISION 67 PAGE 280 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Potential Loss Threshold: I16. RCS leakage > 50 gpm inside the drywell CNS Basis: RCS leakage inside the drywell Is normally determined by monitoring drywell equipment and floor drain sump pump out rates. This method of monitoring leakage may be isolated as part of the drywell isolation, and thus, may be unavailable. If primary system leak rate information ls unavaUable, other indicators of RCS leakage should be used (Reference 1-6). Inventory loss events, such as a stuck open SRV, should not be considered when referring to "RCS leakage" because they are not Indications of a break, which could propagate. CNS Basis Reference(s):

1. System Operating Procedure 2.2.27, Equipment,, Floor, and Chemical Drain System.

2. Alarm Procedure 2.3_9-4-2, Panel 9 Annunciator 9-4-2, B-1/B-2.

3. Surveillance Procedure 6.LOG.601, Daily Surveillance Log - Modes 1, 2, and 3.

4. Technical Specifications LCO 3.4.4, RCS Operational Leakage.

5. Technical Specifications LCO 3.4.5, RCS Leakage Detection Instrumentation.

6. USAR Section X-14.0, Equipment and Floor Drainage Systems.

NEI 99-01 Basis: This threshold is based on leakage set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak; however, break propagation leading to significantly larger loss of inventory is possible. If primary system leak rate information Is unavailable, other indicators of RCS leakage should be used. PROCEDURE 5.7.1 REVISION 67 PAGE 281 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Potential Loss Threshold:

17. Unisolable primary system discharge outside Primary Containment as indicated by exceeding any Secondary Containment Maximum Normal Operating temperature or radiation value (EOP-SA Tables 9 and :1,.0)

CNS Basis: The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of unisolable primary system leakage outside the Primary Containment. The Maximum Normal Operating values define this RCS threshold because they signify the onset of abnormal system operation. When parameters reach this level, equipment failure or mis-operation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-SA, Secondary Containment Control, Tables 9 and 10 (Reference 1) (see below). In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Secondary Containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g., room flooding, high area temperatures, reports of steam in the Secondary Containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the Secondary Containment. As indicated by Note 5 in EOP-SA, Table 10, RP surveys and ARM teledosimetry system may be used for these indications. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 282 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] EOP-SA Table 9 - Secondary Containment Temperatures 9 SECONDARY CONTAINMENT TEMPERATURES SPDS16 Maximum Normal Operating Value Maximum Safe Operating Value 8.cfla 8 )t Iem~ S:tLiti;;b 81allllfld 8raa. 1!alue (~l 8ciual 1!alue NE Quad RCIC-TS-77A NE Quad 195 RCIC-TS-77C SE Quad RWCU-TS-117F SE Quad 195 NW Quad RHR-TS-99C NW Quad 195 SW Quad RHR-TS-99G SW Quad and HPCI-TS-105B and 195 HPCI Room HPCI-TS-105D HPCIRoom 1001' El. 1001' El. 976' El. RWCU-TS-117B 976'EI. 195 958' El. 958' El. RHR-TS-99A RHR-TS-99E 903' El. MS-TS-126A 903' El. 195 and MS-TS-126C and 931' El. RWCU-TS-117E 931' El. RWCU-TS-117A H PCI-TS-105A ( continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 283 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] EOP-SA Table 10 - Secondary Containment Radiation Levels 10 SECONDARY CONTAINMENT RADIATION LEVELS SPDS15 0 Maxunum Safe Maximum Normal Operating Value Operabng Value Ara.a 8i 8BM 81aaned Ba llfl (rnR/hc) Ama :ilalufl (rnBltlc) 8i.11.ial :ilalua FUEL POOL AREA RMA-RA-1 100-1c6 1001' El. 1000 FUEL POOL AREA RMA-RA-2 .01 - 100 1001' El. RWCU PRECOAT AREA RMA-RA-4 0.1 -1000 958' El. RWCU SLUDGE AND DECANT PUMP AREA RMA-RA-5 0.1 - 1000 931' El. 1000 CRD HYDRAULIC EQUIP AREA (SOUTH) RMA-RA-8 .01 -100 903' El . CRD HYDRAULIC EQUIP AREA (NORTH) RMA-RA-9 .01 - 100 HPCI PUMP ROOM RMA-RA-10 .01 -100 HPCI Room RHR PUMP ROOM, (SOUTHWEST) RMA-RA-11 .01 -100 SW Quad 1000 TORUS HPV AREA (SOUTHWESD RMA-RA-27 1.0 -10000 SW Torus RHR PUMP ROOM, (NORTHWEST) RMA-RA-12 .01 -100 NW Quad 1000 RCIC/CORE SPRAY PUMP ROOM, (NORTHEAST) RMA-RA-13 .01 - 100 NE Quad 1000 CORE SPRAY PUMP ROOM, (SOUTHEAST) RMA-RA-14 .01 - 100 SE Quad 1000 NOTES Area radiation levels can be monrt:ored by RP surveys or ARM teledos1matry system CNS Basis Reference(s):

1. EOP-SA, Secondary Containment Control.

(continued on next page) PROCEDURE 5. 7 .1 RE\/ISION 67 PAGE 284 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: Potential loss of RCS based on primary system leakage outside the Primary Containment is determined from temperature or area radiation Maximum Normal Operating values (EOP-5A, Tables 9 and 10) in the areas of the main steam line tunnel, main turbine generator, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside Primary Containment. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage warrant an Alert classification. An unisolable leak which is indicated by a high alarm setpoint escalates to a Site Area Emergency when combined with Containment Barrier Loss Threshold 20 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded. PROCEDURE 5.7.1 REVISION 67 PAGE 285 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: D. ERO Degradation Threat: Loss Threshold: I13. Emergency RPV Depressurlzation is required CNS Basis: Plant symptoms requiring Emergency RPV Depressurization per the EOPs are indicative of a loss of the RCS barrier. If Emergency RPV depressurization is required, the plant Operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary. Revision 3 of the BWROG EPG allows for limiting Reactor Pressure Vessel (RPV) depressurization by reclosing the Safety Relief Valves (SRVs) under certain conditions. This strategy is intended to prolong operation of steam-driven water injection sources required for adequate core cooling (e.g., RCIC System and/or HPCI) following an extended loss of AC power, and thus maintain the core cooling safety function. When the requirement to enact Emergency Depressurizatlon occurs during such an event, Operators will determine if RPV depressurization will result in a loss of RCIC/HPCI, and, if so, terminate depressurization while maintaining RPV pressure as low as practicable. There is no effect on the fission product barrier threshold intent. The relationship between the operationally significant action and the Reactor Coolant System (RCS) barrier status is unchanged, i.e., performing an Emergency RPV Depressurization per EOPs remains indicative of a loss of the RCS barrier, even though the SRVs may be reclosed. RCS mass has been lost to the wetwell and subsequent depressurizatlons may be required (i.e., the ability of the RCS pressure boundary to serve as an effective barrier to a release of fission products has been diminished). In conclusion, the threshold basis of this EAL remains in effect once the Emergency Depressurlzation starts, even though Operators may reclose the SRVs before the full depressurization is complete (Reference 1, 2, 3, 6, 9, and 10). (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 286 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] CNS Basis Reference(s):

1. EOP-lA, RPV Control.

2. EOP-2A, Emergency RPV Depressurization/Steam Cooling.

3. EOP-2B, RPV Flooding.

4. EOP-3A, Primary Containment Control.

5. EOP-5A, Secondary Containment Control, Radioactivity Release Control.

6. EOP-6A, RPV Pressure (Failure to Scram).

7. EOP 6B, Emergency RPV Depressurization (Failure to Scram).

8. EOP-7A, RPV Control (Failure-to-Scram).

9. EOP 7B, RPV Flooding (Failure to Scram).

10. EPFAQ 2015-003.

NEI 99-01 Basis: Plant symptoms requiring Emergency RPV Depressurization are specified in the EOPs (Reference 1, 2, 4, 5, 8) and are Indicative of a loss of the RCS barrier. If Emergency RPV depressurization is required, the plant Operators are directed to open safety relief valves (SRVs) (Reference 2, 7). Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary. PROCEDURE 5.7.1 REVISION 67 PAGE 287 OF 342

, ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: E. Rad Degradation Threat: Loss Threshold: I14. Drywell radiation monitor (RMA-RM-40A/B) > 2.40E+02 Rem/hr CNS Basis: EPIP 5. 7 .17 .1, Dose Assessment (Manual), Core Damage Estimation attachment provides a method of calculating percent fuel clad damage and fuel melt based on drywell radiation. Under LOCA conditions, a reading of 2.44E+6 Rem/hr corresponds to 100% core melt on RMA-RM-40A/B. A value of 2.44E+2 Rem/hr (rounded to 2.40E+02 Rem/hr) yields 0.1 % fuel clad damage using this method. This amount of clad damage is approximately the equivalent of Technical Specification coolant activity discharged uniformly throughout the Primary Containment (Reference 1). CNS Basis Reference(s):

1. EPIP 5. 7.17.1, Dose Assessment (Manual).

NEI 99-01 Basis: The 2.40E+02 Rem/hr value indicates the release of reactor coolant to the Primary Containment. The reading was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere. This reading is less than that specified for Fuel Clad barrier Loss Threshold 2. Thus, this threshold would be indicative of a RCS leak only. If radiation monitor reading increased to that value specified by Fuel Clad Barrier threshold, fuel damage would also be indicated. There is no Potential Loss threshold associated with this item. PROCEDURE 5.7.1 REVISION 67 PAGE 288 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: F. Judgment Degradation Threat: Loss Threshold:

15. Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier CNS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining If the RCS 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 2 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 off-site monitoring results.
Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CNS Basis Reference(s): None NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier Is lost. In addition, the inability to monitor the barrier should also be considered in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost. PROCEDURE 5.7.1 REVISION 67 PAGE 289 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRJERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Reactor Coolant System Category: F. Judgment Degradation Threat: .Potential Loss Threshold:

18. Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier CNS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if RCS 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 2 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 off-site monitoring results.
Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CNS Basis Reference(s): None NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is potentially lost. In addition, the inability to monitor the barrier should also be considered in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost. PROCEDURE 5. 7.1 REVISION 67 PAGE 290 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: A. RPV Level Degradation Threat: Potential Loss Threshold: I2s. SAG 1 entry is required CNS Basis: EOP-lA, EOP-2B, EOP-7A, and EOP-7B specify entry into SAG 1 when it is determined that core damage is occurring due to loss of core cooling (Reference 1). SAG entry signifies the need to implement severe accident mitigation actions. The EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. SAG 1 entry is required during Non-Failure-to-Scram events when any of the following conditions exist (Reference 1):

RPV water level cannot be restored and maintained above -183 inches (MSCRWL, EOP-lA) (Reference 2, 6).
RPV water level cannot be restored and maintained at or above -209 inches (elevation of the jet pump suction) and no core spray subsystem flow can be restored and maintained~ 4,750 gpm (design core spray flow, EOP-lA)

(Reference 2, 7). SAG 1 entry is required during Failure-to-Scram events when RPV water level cannot be restored and maintained > -183 inches and core steam flow cannot be restored and maintained > 800,000 lbm/hr (Reference 3, 6). The specified steam flow is the Minimum Core Steam Flow (MCSF). The MCSF is the lowest core steam flow sufficient to preclude any clad temperature from exceeding 1500°F even if reactor core is not completely covered (Reference 6). The MCSF is only applicable in failure-to-scram events because reactor power must be well above the decay heat generation rate for steam production to equal the MCSF. (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 291 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Whether or not RPV water level can be determined, SAG 1 entry is also required when core damage ls occurring due to loss of core cooling (EOP-lA, EOP-2B, EOP-7A, or EOP-7B) (Reference 2, 3, 4, 5). If RPV water level cannot be determined, the absence of core damage indications may be the only means of determining if adequate core cooling is being maintained. If RPV water level can be determined, restoration of RPV water level to above -183 inches and restoration of core spray cooling requirements may not occur in a timely manner. If indications of core damage occur while RPV injection is being restored, entry to SAG 1 is appropriate even if the required water levels and spray cooling flow are eventually achieved. The above EOP conditions represent a potential core melt sequence which could lead to RPV failure and increased potential for containment failure. This threshold is also a Loss of the Fuel Clad barrier (FC Loss 1). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCS Loss 7). SAG entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. The EALs are aligned to the updated guidance of EPG/SAG Revision 3 such that the plant Operator can readily implement both when needed (Reference 8). CNS Basis Reference( s):

1. AMP-TBD00 PSTG/SATG Technical Bases, RC/L, Contingency #1, #4, #5.

2. EOP-lA, RPV Control.

3. EOP-7A, RPV Level (Failure-to-Scram).

4. EOP-2B, RPV Flooding.

5. EOP-7B, RPV Flooding (Failure-to-Scram).

6. NEDC 97-090J.

7. NEDC 97-089.

8. EPFAQ 2015-004.

(continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 292 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The potential loss requirement for SAG 1 indicates adequate core cooling cannot be established and maintained and that core melt Is possible. Entry Into the SAG 1 is a logical escalation in response to the inability to maintain adequate core cooling. EOPs direct SAG 1 entry when it is determined that core damage is occurring due to loss of core cooling (Reference 1): The condition in this potential loss threshold represents a potential core melt sequence which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with Reactor Vessel water level "Loss" thresholds in the Fuel Clad and RCS barrier columns, this threshold will result In the declaration of a General Emergency - loss of two barriers and the potential loss of a third. PROCEDURE 5.7.1 REVISION 67 PAGE 293 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: B. PC Pressure/Temperature Degradation Threat: Loss Threshold:

19. PC pressure rise followed by a rapid unexplained drop in PC pressure CNS Basis:

None CNS Basis Reference(s): None NEI 99-01 Basis: Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase from a high energy line break indicates a loss of containment integrity. This indicator relies on Operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition. PROCEDURE 5.7.1 REVISION 67 PAGE 294 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Containment Category: B. PC Pressure/Temperature Degradation Threat: Loss Threshold: I20. PC pressure response not consistent with LOCA conditions CNS Basis: Analysis of the Primary Containment response to a postulated OBA LOCA event gives a peak drywell pressure of 54.4 psig and a peak drywell temperature of 301.4°F. These peak values were obtained for the power/flow point of 1020/oP/75°/oF (MELLL point). Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60% to 80% of the analyzed rate. The unexpected response is important because it is the indicator for a containment bypass condition (Reference 1). As stated in the NEI 99-01 basis, the anticipated response to a LOCA is that the Primary Containment pressure would increase. The failure of torus to drywell vacuum breaker(s) could cause peak Primary Containment pressure to be higher than the analyzed peak Primary Containment pressure, but this condition is addressed by the potential containment failure Criteria 26. As such, violation of the pressure suppression pressure curve (PSP) does not constitute a loss of the Primary Containment. CNS Basis Reference(s):

1. USAR Section XIV-6.3. 7.

NEI 99-01 Basis: Primary Containment pressure should increase initially as a result of mass and energy release into containment from a LOCA. Thus, Primary Containment pressure not initially increasing under these conditions indicates a loss of containment integrity. This indicator relies on Operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition. PROCEDURE 5. 7 .1 REVISION 67 PAGE 295 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: B. PC Pressure/Temperature Degradation Threat: Potential Loss Threshold: I26. PC pressure > 56 psig and rising CNS Basis: The Primary Containment internal design pressure is 56 psig (Reference 1). If this threshold is exceeded, a challenge to the Primary Containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Primary Containment barrier even if a containment breach has not occurred. CNS Basis Reference(s):

1. USAR Table V-2-1.

NEI 99-01 Basis: The 56 psig for Potential Loss of containment is based on the Primary Containment design pressure. PROCEDURE 5. 7.1 REVISION 67 PAGE 296 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: B. PC Pressure/Temperature Degradation Threat: Potential Loss Threshold:

27. Deflagration concentrations exist inside PC

    * ~  6% H2 in drywell or torus (or cannot be determined)

AND

    *  ~ 5% 02 In drywell or torus or cannot be determined)

CNS Basis: Deflagration (explosive) mixtures in the Primary Containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to Primary Containment Integrity. Except for brief periods during plant startup and shutdown, oxygen concentration In the Primary Containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen) and readily recognizable because 6% hydrogen is well above the EOP-3, Primary Containment Control, entry condition (Reference 2, 3). Since the EOPs/SAGs require deflagration concentration actions to be performed when hydrogen and oxygen concentrations cannot be determined, the phrase has been added to the* meaning of explosive mixtures. The minimum global deflagration hydrogen/oxygen concentrations (6% and 5%, respectively) require intentional Primary Containment venting, which is defined to be a Loss of Containment (PC Loss 22). (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 297 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Drywell and suppression chamber atmosphere is monitored for H2 and 02 by a divisionally separated H2/O2 Monitoring System. The system consists of two H2/O2 analyzers (PC-AN-H2/O2I and PC-AN-H2/O211), two remote process panels (PC-CS-H2/O2I and PC-CS-H2/O211), two H2 recorders (PC-R-H2I and PC-R-H211), two 02 recorders (PC-R-O21 and PC-R-O211), an 02 digital indicator (PC-1-1), associated control switches and sample stream indicating lights. H2/O2 analyzers are located in the Reactor Building at 976', remote process panels are located in the Cable Spreading Room, recorders are located on VBD-Pl and VBD-P2, the 02 digital indicator and sample stream lights are located on VBD-H. Div 2 is normally In service providing 02 concentration on VBD-H and H2 and 02 concentrations on PMIS (Reference 1). CNS Basis Reference(s):

1. System Operating Procedure 2.2.60.1, Containment H2/O2 Monitoring System.

2. BWROG EPG/SAG Revision 2, Sections PC/G.

3. EOP-3A, Primary Containment Control.

NEI 99-01 Basis: BWRs specifically define the limits associated with explosive (deflagration) mixtures in terms of deflagration concentrations of hydrogen and oxygen. For Mk 1/11 containments, the deflagration limits are "6% hydrogen and 5% oxygen in the drywell or suppression chamber".

PROCEDURE 5. 7 .1 REVISION 67 PAGE 298 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: B. PC Pressure/Temperature Degradation Threat: Potential Loss Threshold:

28. Average torus water temperature and RPV pressure cannot be maintained within the Heat Capacity Temperature Limit (EOP/SAG Graph 7)

CNS Basis: This threshold is met when EOP-3, Primary Containment Control, Step SP/T-5 is reached (Reference 1). CNS Basis Reference(s):

1. EOP-3A, Primary Containment Control.

N EI 99-01 Basis: The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized; or
Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. PROCEDURE 5.7.1 REVISION 67 PAGE 299 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: C. Isolation Degradation Threat: Loss Threshold:

21. Failure of all valves in any one line to close AND Direct downstream pathway to the environment exists after PC isolation signal CNS Basis:

This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of Primary Containment integrity. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main steam line, HPCI steam line, or RCIC steam line breaks, unisolable RWCU System breaks, and unisolable containment atmosphere vent paths. If main condenser is available with an unisolable main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of a non-isolable release path to the environment. These minor releases are assessed using the Category A, Abnormal Rad Release/Rad Effluent, EALs. The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. However, the expectation is that assessment, classification, and declaration of an emergency condition be made within 15 minutes after initial availability of the indication of a breach of Primary Containment integrity. If Operator actions from the Control Room are successful, this threshold is not applicable. Credit is not given for Operator actions taken in-plant (outside the Control Room) to isolate the breach. ( continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 300 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] EOP-3A, Primary Containment Control, Step PC/P-6, may specify Primary Containment venting and intentional bypassing of the containment isolation valve logic, even if off-site radioactivity release rate limits are exceeded (Reference 1). - Under these conditions with a valid containment isolation signal, the containment barrier should be considered lost under Criteria 22 for Intentional PC venting per EOPs. CNS Basis Reference(s):

1. EOP-3A, Primary Containment Control.

NEI 99-01 Basis: These thresholds address incomplete containment isolation that allows direct release to the environment. The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems. The existence of an in-line charcoal filter does not make a release path indirect since the filter is not effective at removing fission product noble gases. Typical filters have an efficiency of 95% to 99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period. PROCEDURE 5. 7 .1 REVISION 67 PAGE 301 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: C. Isolation Degradation Threat: Loss Threshold: I22. Intentional PC venting per EOPs CNS Basis: EOP-3A, Primary Containment Control, Steps PC/P-3 and PC/P-8, may specify Primary Containment venting and intentional bypassing of the containment isolation valve logic, even if off-site radioactivjty release rate limits are exceeded (Reference 1). The threshold is met when the Operator begins venting the Primary Containment in accordance with EOP-3A, not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in EOP-3A, Step PC/P-1, to control Primary Containment pressure below the Primary Containment high pressure scram setpoint does not meet this threshold because such action is only permitted if off-site radioactivity release rates will remain below ODAM limits. CNS Basis Reference(s):

1. EOP-3A, Primary Containment Control.

NEI 99-01 Basis: The EOPs may direct containment isolation valve logic(s) to be Intentionally bypassed, regardless of radioactivity release rates. Under these conditions with a valid containment isolation signal, the containment should also be considered lost if containment venting is actually performed. Intentional venting of Primary Containment for Primary Containment pressure or combustible gas control per EOPs to the Secondary Containment and/or the environment is considered a loss of containment. Containment venting for pressure when not in an accident situation should not be considered. PROCEDURE 5. 7 .1 REVISION 67 PAGE 302 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: C. Isolation Degradation Threat: Loss Threshold:

23. Unisolable primary system discharge outside PC as indicated by exceeding any Secondary Containment Maximum Safe Operating temperature or radiation value (EOP-SA, Tables 9 and 10)

CNS Basis: The Maximum Safe Operating values define this Primary Containment barrier threshold because they are indicative of problems in the Secondary Containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside Primary Containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-SA, Secondary Containment Control, Tables 9 and 10 (see below). In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Secondary Containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g., room flooding, high area temperatures, reports of steam in the Secondary Containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the Secondary Containment. As indicated by NOTE 5 in EOP-SA, Table 10, RP surveys and ARM Teledosimetry System may be used for these indications. (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 303 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] EOP-SA Table 9 - Secondary Containment Temperatures 9 SECONDARY CONTAINMENT TEMPERATURES SPDS16 Maximum Normal Operating Value Maximum Safe Operating Value Acaa. 8 ~ Iamc Switcb 81acmed Araa. ~lua C~} Actual ~alue NE Quad RCIC-TS-77A NE Quad 195 RCIC-TS-77C SE Quad RWCU-TS-117F SE Quad 195 NW Quad RHR-TS-99C NW Quad 195 SW Quad RHR-TS-99G SW Quad and HPCI-TS-105B and 195 HPCI Room HPCI-TS-105D HPCIRoom 1001'EI. 1001' El. 976'EI. RWCU-TS-117B 976' El. 195 958' El. 958'EI. RHR-TS-99A RHR-TS-99E 903' El. MS-TS-126A 903' El. 195 and MS-TS-126C and 931' El. RWCU-TS-117E 931' El. RWCU-TS-117A HPCI-TS-105A (continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 304 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] EOP-SA Table 10 - Secondary Containment Radiation Levels 10 SECONDARY CONTAINMENT RADIATION LEVELS SPDS15 0 M8XITTI um Safe Maximum Normal Operabng Value Operating Value Ar.a.a 8 :,:: 8BM 81aan~ Ba ge (mR!h[) Amil 1!:alue (mR/h[) 81.wal 1!alu1a FUEL POOL AREA RMA-RA-1 100 -1a6 1001' El. 1000 FUEL POOL AREA RMA-RA-2 .01 -100 1001' El. RWCU PRECOAT AREA RMA-RA-4 0.1 -1000 958' El. RWCU SLUDGE AND DECANT PUMP AREA RMA-RA-5 0.1 -1000 931' El. 1000 CRD HYDRAULIC EQUIP AREA (SOUTH) RMA-RA-8 .01 -100 903' El. CRD HYDRAULIC EQUIP AREA (NORTH) RMA-RA-9 .01 -100 HPCI PUMP ROOM RMA-RA-10 .01 -100 HPCI Room RHR PUMP ROOM, (SOUTHWEST) RMA-RA-11 .01 -100 SW Quad 1000 TORUS HPV AREA (SOUTHWESD RMA-RA-27 1.0 -10000 SW Torus RHR PUMP ROOM, (NORTHWEST) RMA-RA-12 .01 -100 NW Quad 1000 RCIC/CORE SPRAY PUMP ROOM, (NORTHEASD RMA-RA-13 .01 -100 NE Quad 1000 CORE SPRAY PUMP ROOM, (SOUTHEAST) RMA-RA-14 .01 -100 SE Quad 1000 NOTES Area radiation levels can be monitored by RP surveys or ARM telados1metry system CNS Basis Reference(s):

1. EOP-SA, Secondary Containment Control.

(continued on next page) PROCEDURE 5. 7 .1 REVISION 67 PAGE 305 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: The presence of area radiation levels or area temperatures above any Maximum Safe Operating value Indicates unlsolable primary system leakage outside the Primary Containment are addressed after a containment isolation. The indicators should be confirmed to be caused by RCS leakage. There is no Potential Loss threshold associated with this item. PROCEDURE 5.7.1 REVISION 67 PAGE 306 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: E. Rad Degradation Threat: Potential Loss Threshold: I29. Drywell radiation monitor (RMA-RM-40A/B) > 5.00E+04 Rem/hr CNS Basis: EPIP 5.7.17.1, Dose Assessment (Manual), Core Damage Estimation attachment provides a method of calculating percent fuel clad damage and fuel melt based on drywell radiation. A reading of 2.44E+6 Rem/hr corresponds to 100% core melt on RMA-RM-40A/B. A value of 4.88E+4 Rem/hr (rounded to 5.00E+04 Rem/hr) yields 20% fuel clad damage using this method (Reference 1). CNS Basis Reference(s):

1. EPIP 5.7.17.1, Dose Assessment (Manual).

NEI 99-01 Basis: 50,000 Rem/hr is a value which indicates significant fuel damage well in excess of that required for loss of RCS and Fuel Clad. A major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a Potential Loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates that such conditions do not exist when the amount of clad damage is

< 20%.

PROCEDURE 5. 7 .Jl REVISION 67 PAGE 307 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Category: F. Judgment Degradation Threat: Loss Threshold:

24. Any condition in the opinion of the Emergency Director that indicates loss of the PC barrier CNS Basis:

The Emergency Director 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 2 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 off-site monitoring results.
Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CNS Basis Reference(s): None (continued on next page) PROCEDURE 5. 7.1 REVISION 67 PAGE 308 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Primary Containment barrier is lost. In addition, the inability to monitor the barrier should also be considered as a factor in Emergency Director judgment that the barrier may be considered lost. The Containment barrier should not be declared lost based on exceeding Technical Specification action statement criteria unless there is an event in-progress requiring mitigation by the Containment barrier. When no event is in-progress (Loss or Potential Loss of either Fuel Clad and/or RCS), the Containment barrier status is addressed by Technical Specifications. PROCEDURE 5. 7 .1 REVISION 67 PAGE 309 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] Barrier: Primary Containment Degradation Threat: Potential Loss Category: F. Judgment Threshold:

30. Any condition in the opinion of the Emergency Director that indicates potential loss of the PC barrier CNS Basis:

The Emergency Director 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 2 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 off-site monitoring results.
Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

CNS Basis Reference(s): None (continued on next page) PROCEDURE 5.7.1 REVISION 67 PAGE 310 OF 342

ATTACHMENT 3 FISSION PRODUCT BARRIERS - INDICATIONS OF LOSS OR POTENTIAL LOSS TECHNICAL BASES [INFORMATION USE] NEI 99-01 Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Primary Containment barrier is potentially lost. In addition, the inability to monitor the barrier should also be considered as a factor in Emergency Director judgment that the barrier may be considered potentially lost. The Containme,nt barrier should not be declared potentially lost based on exceeding Technical Specification action statement criteria unless there is an event in-progress requiring mitigation by the Containment barrier. When no event ls in-progress (Loss or Potential Loss of either Fuel Clad and/or RCS), the Containment barrier status is addressed by Technical Specifications. PROCEDURE 5.7.1 REVISION 67 PAGE 311 OF 342

ATTACHMENT 4 EAL CLASSIFICATION MATRIX [INFORMATION USE] ATTACHMENT 4 EAL Q.ASSIFICATION MATRIX [INFORMATION USE] The following information defined in Attachment 1, EAL Scheme Explanation and Ratlonale, and contained in Attachment 2, Emergency Action Level Technical Bases, will be contained in the EAL Classification Matrix (Matrix or EAL Matrix):

EAL Identifier.
Mode Applicability.
EAL.

The Matrix will also display the tables and notes from Attachments 2 and 3 applicable to the EALs. These items may be reformatted, arranged, and consolidated as required to facilitate use of the Matrix. The EALs will be arranged by Emergency Class left to right, greatest to least, then by Category and subcategory top to bottom, and finally by EAL identifier top to bottom where required. These Matrices will be controlled per this attachment. The information specified above will be word for word from Attachment 2 but may be formatted differently using different font sizes or color backgrounds to assist the visual presentation. Each Matrix will contain a Revision data box that will list the current matrix revision number based on the information below: EAL Classification Matrix Revision Data: Procedure EAL Classification Matrix Revision Number EPIP 5.7.1, Attachment 4 Revision 19 It is not necessary that the Matrix revision number be revised with each revision of this procedure. However, if the Matrix is revised or information specified above (EALs, Notes, or Tables) are revised in Attachment 2, then Attachment 4 and the matrix must be revised to reflect the revised information. Each controlled copy of the matrix will be labeled with the facility and copy number of the specific matrix card according to EPDG#2, Attachment F-5. Matrices that are not so labeled are uncontrolled and should be checked to verify the proper revision prior to use. Matrix distribution will be made to following locations in quantities specified in EPDG #2, Attachment F-5. PROCEDURE 5.7.1 REVISION 67 PAGE 312 OF 342

ATTACHMENT 4 EAL CLASSIFICATION MATRIX [INFORMATION USE] EAL Classification Matrix Locations:

1. Control Room.

2. Simulator.

3. Emergency Operations Facility.

4. Technical Support Center.

5. Joint Information Center.

6. Emergency Preparedness Office.

PROCEDURE 5.7.1 REVISION 67 PAGE 313 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] ATTACHMENT 5 EAL GROUPS, CATEGORIES, ANO SUBCATEGORIES [INFORMATION USE] EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: A - Abnormal Rad Release/Rad 1 - Off-Site Rad Conditions Effluent 2 - On-Site Rad Conditions and Spent Fuel Pool Events H - Hazards And Other Conditions 1 - Natural or Destructive Phenomena Affecting Plant Safety 2 - Fire or Explosion 3 - Hazardous Gas 4 - Security 5 - Control Room Evacuation 6 - Judgment E - ISFSI None MODES l, 2, or 3: S - System Malfunction 1 - Loss of AC Power 2 - ATWS/Criticality 3 - Inability to Reach Shutdown Conditions 4 - Instrumentation 5 - Fuel Clad Degradation 6 - RCS Leakage 7 - Loss of DC Power 8 - Communications F - Fission Product Barrier None Degradation MODES 4, 5, or DEF: C - Cold Shutdown/Refuel System 1 - Loss of AC Power Malfunction 2 - RPV Level 3 - RCS Temperature 4 - Communications 5 - Inadvertent Criticality 6 - Loss of DC Power PROCEDURE 5.7.1 REVISION 67 PAGE 314 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] CATEGORY A - ABNORMAL RAD RELEASE/RAD EFFLUENT EAL Group: ANY (EALs in this category are applicable to any plant condition) Many EALS are based on actual or potential degradation of fission product barriers because of the elevated potentlal for off-site 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, off-site radiological conditions may result which require off-site 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:

7. OFF-SITE RAD CONDillONS Direct indication of effluent Radiation Monitoring Systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits.

Projected off-site doses, actual off-site field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. ON-SITE RAD CONDffiONS AND SPENT FUEL POOL EVENTS Sustained general area radiation levels in excess of those Indicating loss of control of radioactive materials or those levels which may preclude access to vital plant areas also warrant emergency classification.

PROCEDURE 5. 7 .1 REVISION 67 PAGE 315 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] CATEGORY H - HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY EAL GROUP: ANY (EALS in this category are applicable to any plant condition) Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety, or personnel safety. The events of this category pertain to the following subcategories:

3. NATURAL OR DESTRUCTIVE PHENOMENA Natural events include earthquakes, tornados, high winds, and high/low river levels that have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. Non-naturally occurring events that can cause damage to plant facilities and include vehicle crashes, missile impacts, internal flooding, etc.

2. FIRE OR EXPLOSION 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 vital equipment.

3. HAZARDOUS GAS Non-naturally occurring events that can cause damage to plant facilities and include toxic, corrosive, asphyxiant, or flammable gas leaks.

4. SECURITY Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

5. CONTROL ROOM EVACUATION Events indicative of loss of Control Room habitability. If Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the Emergency Response Facilities.

PROCEDURE 5. 7 .1 REVISION 67 PAGE 316 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE]

6. JUDGMENT The EALs defined in other categories specify the pre-determined 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 Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

PROCEDURE 5.7.1 REVISION 67 PAGE 317 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] CATEGORY E - ISFSI EAL GROUP: ANY (the EAL in this category is applicable to any plant condition) 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 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. Formal off-site planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety. 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. This includes classification based on a loaded fuel storage cask confinement boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage. A security event that leads to a potential loss of level of safety of the ISFSI is a classifiable event under Security Category EAL HU4.1. Minor surface damage that does not affect storage cask boundary is excluded from the scope of these EALs. PROCEDURE 5.7.1 REVISION 67 PAGE 318 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] CATEGORY S - SYSTEM MALFUNCTION EAL GROUP: MODES 1, 2, OR 3 Numerous system-related equipment failure events that warrant emergency classificatlon 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:

7. LOSS OF 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.

8. ATWS/CRIDCALITY Events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events.

For EAL classification, however, ATWS is Intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to Fuel Clad, RCS, and Containment integrity. Inadvertent criticallties pose potential personnel safety hazards as well being indicative of losses of reactivity control.

9. INABILITY TO REACH SHUTDOWN CONDITIONS One EAL falls into this subcategory. It is related to the failure of the plant to be brought to the required plant operating condition required by Technical Specifications If a limiting condition for operation (LCO) is not met.

10. INSTRUMENTATION Certain events that degrade plant Operator ability to effectively assess plant conditions within the plant warrant emergency classification. Loss of annunciators or indicators is In this subcategory.

PROCEDURE 5.7.1 REVISION 67 PAGE 319 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE]

11. FUEL CLAD DEGRADATION During normal operation, reactor coolant fission product activity is very low.

Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (1 % clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

12. RCS LEAKAGE The Reactor Vessel provides a volume for the coolant that covers the reactor core. The Reactor Vessel and associated pressure piping (Reactor Coolant System) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail.

Excessive RCS leakage greater than Technical Specification limits are utilized to indicate potential pipe cracks that may propagate to an extent threatening fuel clad, RCS, and containment integrity.

13. LOSS OF DC POWER Loss of vital critical DC 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 125 VDC power sources.

14. COMMUNICATIONS Certain events that degrade plant Operator abillty to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

PROCEDURE 5. 7 .1 REVISION 67 PAGE 320 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] CATEGORY F - FISSION PRODUCT BARRIER DEGRADATION© 4 EAL GROUP: MODE 1, 2, OR 3 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 zircaloy fuel bundle tubes that contain the fuel pellets. B. Reactor Coolant System (RCS): The RCS barrier is the Reactor Coolant System pressure boundary and includes the reactor vessel and all Reactor Coolant System piping up to the isolation valves. C. Primary Containment (PC): The Primary Containment barrier includes the , drywell, the wetwell (torus), their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. 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 3). "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:

Unusual Event: Any loss or any potential loss of Primary Containment.
Alert: Any loss or any potential loss of either Fuel Clad or RCS.
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.

PROCEDURE 5. 7.1 REVISION 67 PAGE 321 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Primary Containment barrier. UE EALs associated with RCS and Fuel Clad barriers are addressed under System Malfunction EALs.
At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if Fuel Clad and RCS barrier "Loss" EALs existed, that, in addition to off-site dose assessments, would require continual assessments of radioactive inventory and containment integrity. Alternatively, if both Fuel Clad and RCS barrier "Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
The ability to escalate to higher emergency classes as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would represent an Increasing risk to public health and safety.
The Primary Containment barrier should not be declared lost or potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in-progress requiring mitigation by the Primary Containment barrier. When no event is in-progress (Loss or Potential Loss of either Fuel Clad and/or RCS), the Primary Containment barrier status is addressed by Technical Specifications.

Determine which combination of the three barriers are lost or have a potential loss and use FUl.1, FAl.1, FSl.1, and FGl.1 to classify the event. Also, an event for multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds is imminent. In this Imminent loss situation, use judgment and classify as If the thresholds are exceeded. PROCEDURE 5. 7 .1 REVISION 67 PAGE 322 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE] CATEGORY C - COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION EAL GROUP: MODES 4, 5, DEF Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out of service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that Instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, Containment Closure, and fuel clad integrity for the applicable operating modes ( 4 - Cold Shutdown, 5 - Refueling, D - Defueled). THE EVENTS OF THIS CATEGORY PERTAIN TO THE FOLLOWING SUBCATEGORIES:

15. LOSS OF AC POWER Loss of emergency plant electrical power can compromise plant safety system operability including Decay Heat Removal and Emergency Core Cooling Systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of on-site and off-site sources for 4160V emergency buses and loss of vital 125 VDC power sources.

16. RPV LEVEL RPV water level is a measure of inventory available to ensure adequate core cooling and, therefore, maintain fuel clad integrity. The RPV provides a volume for the coolant that covers the reactor core. The RPV 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.

17. RCS TEMPERATURE Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

PROCEDURE 5.7.1 REVISION 67 PAGE 323 OF 342

ATTACHMENT 5 EAL GROUPS, CATEGORIES, AND SUBCATEGORIES [INFORMATION USE]

18. COMMUNICATIONS Certain events that degrade plant Operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

19. INADVERTENT CRITICALITY Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.

20. LOSS OF DC POWER Loss of vital critical DC 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 125 VDC power sources. PROCEDURE 5. 7.1 REVISION 67 PAGE 324 OF 342

ATTACHMENT 6 EAL DEFINmONS AND ACRONYMS [INFORMATION USE] ATTAD-IMENT 6 EAL DEFINmONS AND ACRONYMS [INFORMATION USE] DEFINITIONS Affecting Safe Shutdown Event in-progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable hot or cold shutdown condition. Plant condition applicability is determined by Technical Specification LCOs in effect. Example 1: Event causes damage that results in entry into an LCO that requires the plant to be placed in hot shutdown. Hot shutdown is achievable, but cold shutdown is not. This event is not "affecting safe shutdown". Example 2: Event causes damage that results in entry into an LCO that requires the plant to be placed in cold shutdown. Hot shutdown is achievable, but cold shutdown is not. This event is "affecting safe shutdown". Alert Events are in-progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. Bomb Refers to an explosive device suspected of having sufficient force to damage plant systems or structures. Civil Disturbance A group of people violently protesting station operations or activities at the site. Confinement Boundary Is the barrier(s) between areas containing radioactive substances and the environment. Containment Closure Is the action taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment Closure requirements are specified in Administrative Procedl!re 0.50.5, Outage Shutdown Safety. PROCEDURE 5. 7 .1 REVISION 67 PAGE 325 OF 342

ATTACHMENT 6 EAL DEFINffiONS AND ACRONYMS [INFORMATION USE] Emergency Action Level (EAL) A pre-determined, site specific, observable threshold for a plant IC that places the plant in a given emergency classification level. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (on-site or off-site); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency classification level. Emergency Classification Level (ECL) One of a minimum set of names or titles established by the NRC for grouping off normal nuclear power plant conditions according to (1) their relative radiological seriousness, and (2) the time-sensitive on-site and off-site radiological emergency preparedness actions necessary to respond to such conditions. The existing radiological emergency classification levels, in ascending order of seriousness, are called:

Notification of Unusual Event (UE).
Alert.
Site Area Emergency (SAE).
General Emergency (GE).

Explosion A rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components. Extortion Is an attempt to cause an action at the station by threat of force. 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 Flooding, as used within the EALs, describes a condition where water is entering a room faster than installed equipment is capable of removal, resulting in a rise of water level within the room. PROCEDURE 5.7.1 REVISION 67 PAGE 326 OF 342

ATTACHMENT 6 EAL DEFINffiONS AND ACRONYMS [INFORMATION USE] General Emergency Events are in-progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile action that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels off-site for more than the immediate site area. Hostage Person(s) held as leverage against the station to ensure demands will be met by the station. Hostile Action An act toward CNS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidates 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 CNS. Non-terrorism-based EALs should be used to address such activities (e.g., 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 Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates the event or condition will occur. Where IMMINENT timeframes are specified, they shall apply. Initiating Condition (IC) One of a pre-determined subset of nuclear power plant conditions where either the potential exists for a radiological emergency or such an emergency has occurred. Inoperable Not able to perform its intended function. PROCEDURE 5.7.1 REVISION 67 PAGE 327 OF 342

ATTACHMENT 6 EAL DEFINITIONS AND ACRONYMS [INFORMATION USE] Intruder Person(s) present in a specified area without authorization. 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. Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Normal Plant Operations Activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from Normal Plant Operations. Operating Mode Applicability The plant mode existing at the start of an event under which a particular EAL is applicable. These modes are defined in Technical Specifications Table 1.1-1. The Defueled or DEF mode referred to in some EALs is the condition where all fuel has been removed from the reactor vessel. Note that the ISFSI EAL has no mode applicability. Projectile An object directed toward CNS that could cause concern for its continued operability, reliability, or personnel safety. Protected Area An area which normally encompasses all controlled areas within the security Protected Area fence. Sabotage Deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may not meet the definition of sabotage until this determination is made by Security Supervision. PROCEDURE 5. 7.1 REVISION 67 PAGE 328 OF 342

ATTACHMENT 6 EAL DEFINffiONS AND ACRONYMS [INFORMATION USE] 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. Significant Transient An unplanned event involving any of the following:

Runback > 25% thermal power.
Electrical load rejection > 25% full electrical load.
Reactor scram.
ECCS injection.
Thermal power oscillations > 10%.

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. Strike Action Work stoppage within the Protected Area by a body of workers to enforce compliance with demands made on CNS. The strike action must threaten to interrupt Normal Plant Operations. Sustained Wind Sustained winds are of a prolonged duration and, therefore, do not include gusts. Sustained winds are not Intermittent or of a transitory nature. Since the inauguration of the Automatic Surface Observation System (ASOS), the National Weather Service has adopted a 2 minute average standard for its sustained wind definition. Unisolable A breach or leak that cannot be promptly isolated. PROCEDURE 5.7.1 REVISION 67 PAGE 329 OF 342

ATTACHMENT 6 EAL DEFINITIONS AND ACRONYMS [INFORMATION USE] Unplanned A parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions. 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 off-site 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 conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment. Visible Damage Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage Is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Examples of damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included. Vital Area Any area, normally within the Protected Area, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. Acronyms AC ..................... Alternating Current ADS ................... Automatic Depressurization System APRM ................. Average Power Range Meter ATWS .....*............ Anticipated Transient Without Scram BIIT ................... Boron Injection Initiation Temperature BWR .................. Boiling Water Reactor PROCEDURE 5.7.1 REVISION 67 PAGE 330 OF 342

ATTACHMENT 6 EAL DEFINffiONS AND ACRONYMS [INFORMATION USE] CCW .................. Component Cooling Water CDE ................... Committed Dose Equivalent CFR .................... Code of Federal Regulations cps .................... Counts per Second CRD ................... Control Rod Drive CS ..................... Core Spray CST ................... Condensate Storage Tank CTMT/CNMT ........ Containment OBA ................... Design Basis Accident DC ..................... Direct Current De min ................ Demineralizer DHRP ................. Decay Heat Removal Pressure DOT ................... Department of Transportation DW .................... Drywell DWSIL. ............... Drywell Spray Initiation Limit EAL .................... Emergency Action Level ECCS ................. Emergency Core Cooling System ECL .................... Emergency Classification Level ED ..................... Emergency Director El. ..................... Elevation EOF ................... Emergency Operations Facility EOP ................... Emergency Operating Procedure EPA .................... Environmental Protection Agency PROCEDURE 5. 7.1 REVISION 67 PAGE 331 OF 342

ATTACHMENT 6 EAL DEFINmONS AND ACRONYMS [INFORMATION USE] EPG ................... Emergency Procedure Guideline EPIP ................... Emergency Plan Implementing Procedure EPRI .................. Electric Power Research Institute ERD ................... Emergency RPV Depressurization ESF .................... Engineered Safety Feature ESW ................... Emergency Service Water FAA .................... Federal Aviation Administration FBI .................... Federal Bureau o,f Investigation FEMA ................. Federal Emergency Management Agency FSAR .................. Final Safety Analysis Report ft ....................... Feet gal ..................... Gallon(s) GE .. : .................. General Emergency GPM ................... Gallons Per Minute HCTL .................. Heat Capacity Temperature Limit HCU ................... Hydraulic Control Unit HOO ................... Headquarters (NRC) Operations Officer HPCI .................. High Pressure Coolant Injection H2 ..................... Hydrogen hr ...................... Hour HX ..................... Heat Exchanger IC ...................... Initiating Condition in ...................... Inch(es) PROCEDURE 5. 7 .1 REVISION 67 PAGE 332 OF 342

ATTACHMENT 6 EAL DEFINITIONS AND ACRONYMS [INFORMATION USE] IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI ................. Independent Spent Fuel Storage Installation Keff ................... Effective Neutron Multiplication Factor lb ....................... Pound(s) LCO ................... Limiting Condition of Operation LER .................... Licensee Event Report LOCA ................. Loss of Coolant Accident LPCI. .................. Low Pressure Coolant Injection LWR ................... Light Water Reactor MDRIR ............... Minimum Debris Retention Injection Rate MDSL ................. Minimum Debris Submergence Level MELLL ................ Maximum Extended Load Line Limit min .................... Minimum, minute mR .................... milliRoentgen mRem ................ milliRem MSCP ................. Minimum Steam Cooling Pressure MSIV .................. Main Steam Isolation Valve MSL ................... Main Steam Line MW .................... Megawatt N/A .................... Not applicable NEI .................... Nuclear Energy Institute NESP .................. National Environmental Studies Project PROCEDURE 5. 7.1 REVISION 67 PAGE 333 OF 342

ATTACHMENT 6 EAL DEFINffiONS AND ACRONYMS [INFORMATION USE] NORAD ............... North American Aerospace Defense Command NPP .................... Nuclear Power Plant NR ..................... Narrow Range NRC ................... Nuclear Regulatory Commission NSSS ................. Nuclear Steam Supply System NU MARC ............. Nuclear Management and Resources Council 02 ...................... Oxygen OBE ................... Operating Basis Earthquake OCA ................... Owner Controlled Area ODCM/ODAM ....... Off-site Dose Calculation (Assessment) Manual ORO ................... Off-site Response Organization PA ..................... Protected Area PAG ................... Protective Action Guideline PC ..................... Primary Containment PCPL .................. Primary Containment Pressure Limit PMIS .................. Plant Management Information System POAH ................. Point of Adding Heat PRA/PSA ............. Probabilistic Risk Assessment/Probabilistic Safety Assessment PRM ................... Process Radiation Monitor psig ................... Pounds per square inch (gauge) PSP .................... Pressure Suppression Pressure PSTG .................. Plant Specific Technical Guidelines R ....................... Roentgen PROCEDURE 5. 7 .1 REVISION 67 PAGE 334 OF 342

ATTACHMENT 6 EAL DEFINmONS AND ACRONYMS [INFORMATION USE] RB ..................... Reactor Building RCC ................... Reactor Control Console RCIC .................. Reactor Core Isolation Cooling RCS ................... Reactor Coolant System rem ................... Roentgen Equivalent Man RETS .................. Radiological Effluent Technical Specifications RHR ................... Residual Heat Removal RPS .................... Reactor Protection System RPV .................... Reactor Pressure Vessel RWCU ................ Reactor Water Cleanup SAG ................... Severe Accident Guideline SFP .................... Spent Fuel Pool SGT ................... Standby Gas Treatment SBO ................... Station Blackout SLC .................... Standby Liquid Control SPDS ................. Safety Parameter Display System SRO ................... Senior Reactor Operator SRV ................... Safety Relief Valve SSE ................... Safe Shutdown Earthquake TAF .................... Top of Active Fuel TEDE .................. Total Effective Dose Equivalent TSC ................... Technic,al Support Center UE ..................... Notification Of Unusual Event PROCEDURE 5. 7 .1 REVISION 67 PAGE 335 OF 342

ATTACHMENT 6 EAL DEFINffiONS AND ACRONYMS [INFORMATION USE] USAR ................. Updated Final Safety Analysis Report WR .................... Wide Range ' ........................ Feet " ........................ Inches % ...................... Percent & ....................... Ampersand ("and") °F ...................... Degrees Fahrenheit > ....................... Greater Than < ....................... Less Than ~ ....................... Greater Than or Equal To ~ ....................... Less Than or Equal To PROCEDURE 5. 7 .1 REVISION 67 PAGE 336 OF 342

ATTACHMENT 7 INFORMATION SHEET [INFORMATION USE] ATTAo-tMENT 7 INFORMATION SHEET [INFORMATION USE]

1. PURPOSE 1.1 Procedure contains the instructions necessary for classification of emergencies consistent with the NRC approved EAL classification scheme. Also, included are the explanations and rational for the scheme, the detailed bases for the EALs, and controls required for the EAL Classification Matrix which is the primary tool used to determine when classification criteria are exceeded.

1.2 Procedure provides the formal set of threshold conditions necessary to classify an event at CNS into one of the four emergency classifications described in NUREG-0654, NEI 99-01, Revision 5, and the CNS Emergency Plan.© 2

2. PRECAUTIONS AND LIMITATIONS 2.1 Assessment, classification, and declaration of an emergency condition shall be completed within 15 minutes after initial availability of indications to plant Operators that an EAL has been exceeded provided that:

2.1.1 Implementation of response actions required to protect public health and safety are not delayed; and, 2.1.2 Any delay in declaration does not deny State and Local authorities opportunity to implement measures necessary to protect public health and safety. 2.2 Classifying and declaration of an emergency is a non-delegable responsibility of Emergency Director. Although additional input in these decisions is encouraged, completion of timely and accurate performance of these activities rests solely with Emergency Director. 2.3 EPIP 5. 7.1 is considered part of CNS Emergency Plari. Proposed changes must be processed per Procedure 0.29.1.

3. REFERENCES 3.1 CODES AND STANDARDS 3.1.1 10CFR50.48, Fire Protection.

PROCEDURE 5. 7 .1 REVISION 67 PAGE 337 OF 342

ATTACHMENT 7 INFORMATION SHEET [INFORMATION USE] 3.1.2 10CFR50. 72, Immediate Notification Requirements for Operating Nuclear Power Reactors. 3.1.3 10CFR72.32, Emergency Plan. 3.1.4 NEI 99-01, Revision 5, Methodology for the Development of Emergency Action Levels, February 2008 (ADAMS Accession Number ML080450149). 3.1.5 NPPD Emergency Plan for CNS. 3.1.6 NUREG-0654, Revision 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants. 3.1.7 NUREG-1022, Event Reporting Guidelines 10CFR50.72 and 50.73, Revision 2. 3.2 PROCEDURES 3.2.1 Instrument Operating Procedure 4.12, Seismic Instrumentation. 3.2.2 Emergency Plan Implementing Procedure 5. 7.1.1, Emergency Classification Process. 3.2.3 Emergency Plan Implementing Procedure 5. 7.2, Emergency Director EPIP. 3.2.4 Emergency Plan Implementing Procedure 5. 7.16, Release Rate Determination. 3.2.5 Emergency Plan Implementing Procedure 5.7.17, CNS-DOSE Assessment. 3.2.6 Emergency Plan Implementing Procedure 5.7.17.1, Dose Assessment (Manual). 3.3 MISCELLANEOUS 3.3.1 ADAMS Accession No. ML100080231, Cooper Nuclear Station - Change to Emergency Action Level Scheme (TAC No. ME0849). Document ID NRC2010008. PROCEDURE 5.7.1 REVISION 67 PAGE 338 OF 342

ATTACHMENT 7 INFORMATION SHEET [INFORMATION USE] 3.3.2 ADAMS Accession No. ML14055A023, Cooper Nuclear Station - Issuance of Amendment Regarding Transition to Risk-Informed, Performance-Based Fire Protection Program in Accordance With 10CFR50.48(c) (TAC No. ME8551). 3.3.3 ADAMS Accession No. ML15271A299, EPFAQ 2015-004, Fission Barrier Matrix Criteria, Date Accepted 01-Jul-15. 3.3.4 RIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events. 3.3.5 © 1 NRC Bulletin 2005-02 (Commitment Number NLS2005080-02), EALs reflect Information provided in Attachment 2 of bulletin. Commitment affects EALs HA4.1, HS4.1, and HG4.1. 3.3.6 © 2 IR 81-01~ (Commitment Number 811217-01-07), Develop Functional Procedure for Accident Classification. Commitment affects Attachment 7, Step 1.2. 3.3.7 © 3 IR 92-014 (Commitment Number 921110-02-03), Revise Guidance Documents for "Low Threshold" Core Damage Events. Commitment affects Attachment 3, Threshold 4, 5, and 6 CNS Basis. 3.3.8 © 4 NRC Document NLS2014035 (Commitment Number NLS2014035-02), Maintain the capability for classifying fuel damage events at the Alert level threshold for CNS at radioactivity levels 300 µCi/g dose equivalent iodine. Commitment affects Attachment 3, Table F-1, on Page 260 and Page 268 Fuel Clad Barrier Criteria 3. Commitment affects Attachment 5, Fission Product Barrier Degradation, Pages 321 through 322. PROCEDURE 5. 7.1 REVISION 67 PAGE 339 OF 342

ATTACHMENT 8 MATRIX BASIS CROSS-REFERENCE [INFORMATION USE] ATTA CHMENT 8 MAT RIX BASI S CROSS - REFERE NCE [IN FORMATION USE] CAT SUB GE SAE ALERT UE AGl.1 39 ASl.l 33 AAl.1 25 AUl.l 16 1 AGl.2 41 ASl.2 35 AAl.2 28 AUl.2 20 AGl.3 43 ASl.3 37 AAl.3 31 AUl.3 23 A AA2.1 51 AU2.1 45 2 AA2.2 53 AU2.2 49 3 AA3.1 55 HAl.1 133 HUl.1 121 HAl.2 137 HUl.2 125 HAl.3 140 HUl.3 127 1 HAl.4 143 HUl.4 129 HAl.5 146 HUl.5 131 HAl.6 148 H HA2.l 155 HU2.l 151 2 HU2.2 153 HA3.1 165 HU3.l 158 3 HU3 .2 163 4 HG4.l 179 HS4.1 177 HA4.l 174 HU4.1 171 5 HS5.1 182 HA5 . 1 181 6 HG6.1 190 HS6.l 188 HA6.l 186 HU6.1 184 E EUl.1 252 1 SGl.1 210 SSl.1 204 SAl.1 198 SUl.1 192 2 SG2.l 224 SS2.1 221 SA2.1 217 SU2.1 216 3 SU3.l 228 4 SS4.l 235 SA4.l 232 SU4.1 229 s SU5.l 239 5 SU5.2 241 6 SU6.l 242 7 SS7.l 245 8 SU8.l 249 PROCEDURE 5. 7 .1 REVISION 67 PAGE 340 OF 342

ATTACHMENT 8 MATRIX BASIS CROSS-REFERENCE [INFORMATION USE] CAT SUB GE SAE ALERT UE F 1 FGl.1 257 FSl .1 255 FAl.1 254 FUl.l 253 1 CAl .1 63 CUl.l 57 CG2.l 87 CS2.1 79 CA2.l 76 CU2.l 69 2 CG2.2 93 CS2.2 81 CU2 .2 71 CS2.3 83 CU2.3 73 C CA3.l 108 CU3.1 101 3 CU3.2 104 4 CU4.1 112 5 CU5.l 115 6 CU6.l 117 PROCEDURE 5. 7 .1 REVISION 67 PAGE 341 OF 342

ATTACHMENT 8 MATRIX BASIS CROSS-REFERENCE [INFORMATION USE] Fission Product Barrier Matrix Reactor Coolant Sys Fuel Clad Barrier Barrier Primary Cont. Barrier Loss Pot. Loss Loss Pot. Loss Loss Pot. Loss A 1 261 8 264 10 274 25 291 11 277 19 294 26 296 B 20 295 27 297 28 299 12 279 16 281 21 300 C 17 282 22 302 23 303 D 13 286 2 267 E 3 268 14 288 29 307 4 269 5 270 6 271 F 7 272 9 273 15 289 18 290 24 308 30 310 PROCEDURE 5. 7.1 REVISION 67 PAGE 342 OF 342

NLS2021001 Page 1 of 9 ENCLOSURE3 Cooper Nuclear Station Emergency Plan Implementing Procedure 5.7.1.1, Emergency Classification Process, Revision 0

COOPER NUCLEAR STATION Operations Manual Emergency Preparedness EMERGENCY PLAN IMPLEMENTING PROCEDURE 5.7.1.1 EMERGENCY CLASSIFICATION PROCESS Level of Use: INFORMATION Quality: QAPD RELATED Effective Date: 12/16/20 Approval Authority: ITR-RDM Procedure Owner: EMERG PREP DRILL SCENARIO COORD PROCEDURE 5.7.1.1 REVISION 0 PAGE 1 OF 8

TABLE OF CONTENTS

1. ENTRY CONDffiONS ............................................................................... 3

2. INffiAL CLASSIFICATION AND DECLARATION ............................................ 3

3. MISSED EAL CLASSIFICATION ................................................................. 5 ATTACHMENT 1 INFORMATION SHEET ................................................... 6 PROCEDURE 5. 7.1.1 REVISION 0 PAGE 2 OF 8

1. ENTRY CONDffiONS 1.1 An Emergency Operation Procedure has been initiated; or 1.2 An unusual occurrence has taken place at or near site.

2. INffiAL CLASSIFICATION AND DECLARATION 2.1 AFTER recognition of off-normal event, THEN ED PERFORM following within 15 minutes:

2.1.1 IF following conditions met:

ED stationed in Control Room (CR).
Potential EAL entry condition exists.
Shift Communicator not in Control Room.

THEN NOTIFY Shift Communicator to come to CR. 2.1.2 ASSESS plant conditions. 2.1.3 IF ED stationed in CR, THEN MAINTAIN oversight of CR team while STE reviews EALs. 2.1.4 IF ED stationed in CR, THEN TURN OVER oversight of CR team to STE. 2.1.5 EVALUATE following applicable EAL Wallchart sections: 2.1.5.1 EAL Matrix from left to right and top to bottom. 2.1.5.2 EAL Category. 2.1.5.3 EAL subcategory. 2.1.5.4 Initiating Condition, as applicable. 2.1.5.5 Mode Applicability bar. 2.1.5.6 Category number criterion. 2.1.5. 7 Applicable notes or tables. 2.1.6 IF event meets an EAL initiating condition and 15-minute time requirement is not challenged, THEN REVIEW EPIP 5.7.1, Attachment 2 or 3. 2.1. 7 DETERMINE applicable Initiating Condition and highest applicable emergency classification level. PROCEDURE 5. 7.1.1 REVISION 0 PAGE 3 OF 8

2.1.8 PERFORM joint discussion with peer checker (CR - STE; EOF - Ops/EOP Advisor, SM, TSC Director). Peer checkers advocate first. 2 .1. 9 DECLARE event. 2.1.10 UPDATE Control Room/EOF team of classification and upgrade criteria. 2.1.11 ENTER EPIP 5.7.2. 2.1.12 DOCUMENT following declaration information as time permits: 2.1.12.1 Time Initiating Condition/EAL threshold is known. 2.1.12.2 Indications of event preceding declaration. 2.1.12.3 Time of declaration. i NOTE - Specific direction for upgrading or terminating an event is given

In EPIP 5. 7.2.

2.1.13 CONTINUE to monitor conditions for potential changes/upgrades to Emergency Classification Level and additional EALs, and REPEAT Steps 2.1.1 through 2.1.12, as necessary, for additional classifications. 2.2 AFTER SM has been relieved of Emergency Director duties, THEN SM INFORM ED of any changing emergency conditions. PROCEDURE 5.7.1.1 REVISION 0 PAGE 4 OF 8

3. MISSED EAL CLASSIFICATION 3.1 IF following conditions met:

Condition existed which met an EAL, but no emergency was declared.
Basis for Emergency Class no longer exists (e.g., condition occurred yesterday but was not caught at time or condition cleared before classification could be made and emergency response is no longer needed).

THEN PERFORM following: 3.1.1 NOTIFY NRC within 1 hour of discovery of undeclared (or misclassified) event per Procedure 2.0.5. 3.1.2 INFORM Emergency Preparedness Manager of details. 3.1.2.1 Emergency Preparedness staff member CONTACT Responsible State and Local Governmental Agencies.

a. PROVIDE details of undeclared event.

PROCEDURE 5.7.1.1 REVISION 0 PAGE 5 OF 8

ATTACHMENT 1 INFORMATION SHEET ATTAO-IMENT 1 INFORMATION SHEET

1. PURPOSE 1.1 Instructions for evaluation, classification, and declaration of an emergency at Cooper Nuclear Station.

2. PRECAUTIONS AND LIMITATIONS 2.1 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 that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following the identification of the appropriate emergency classification level.

2.2 Shift Manager/Emergency Director and Shift Technical Engineer EAL reviews should be performed in series using independent copies of the EAL Matrices except when performing In series would challenge the 15-minute declaration requirement. 2.3 EOF Emergency Director and EOF OPS/EOP Advisor EAL reviews should be performed in parallel using independent copies of the EAL Matrices. 2.4 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS 2.4.1 When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs and associated Initiating Conditions. The highest appllcable Emergency Classification Level identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared.
If two Alert EALs are met, an Alert should be declared.

2.5 The primary tool for determining the emergency classifi_cation level is the EAL Classification Wallcharts. The EAL technical bases should be referenced in order to obtain additional information concerning the EALs under classification consideration. This review shall not challenge the 15-minute declaration time requirement. PROCEDURE 5. 7 .1.1 REVISION 0 PAGE 6 OF 8

ATTACHMENT 1 INFORMATION SHEET

3. RESPONSIBILmES 3.1 SHIFT MANAGER/EMERGENCY DIRECTOR 3.1.1 Evaluate, classify, and declare emergencies prior to EOF becoming operational and assuming the Emergency Director responsibilities.

3.1.2 During determination of emergency classifications, maintain oversight of Control Room activities while the Shift Technical Engineer (STE) performs Emergency Action Level (EAL) review. 3.2 SHIFT TECHNICAL ENGINEER 3.2.1 Independently review the EAL Matrices and provide input to the Shift Manager/Emergency Director. 3.2.2 During determination of emergency classifications, maintain oversight of Control Room activities while the Shift Manager performs EAL review. 3.3 EOF EMERGENCY DIRECTOR 3.3.1 Evaluate, classify, and declare emergencies after the EOF becomes operational and ED responsibilities have been assumed. 3.4 EOF OPERATIONS/EDP ADVISOR 3.4.1 Independently review the EAL Matrices and provide input to the Emergency Director.

4. RECORDS 4.1 No quality records generated by this procedure.

5. REFERENCES 5.1 CODES AND STANDARDS 5.1.1 10CFR50.

5.1.2 NE! 99-02, Regulatory Assessment Performance Indicator Guideline. 5.1.3 NPPD Emergency Plan for CNS. PROCEDURE 5.7.1.1 REVISION 0 PAGE 7 OF 8

ATTACHMENT 1 INFORMATION SHEET 5.1.4 NUREG-0654/FEMA-REP-1, Rev. 1. 5.1.5 NUREG-0654/FEMA-REP-1, Rev. 1, Supplement 3. 5.2 PROCEDURES 5.2.1 Emergency Plan Implementing Procedure 5. 7.1, Emergency Action Level Classification. PROCEDURE 5. 7 .1.1 REVISION 0 PAGE 8 OF 8

NLS2021001 Page 1 of78 ENCLOSURE4 Cooper Nuclear Station On-Shift Staffing Analysis, Revision 2

CNS ON-SHIFT STAFFING ANALYIS REPORT Nebraska Public Power District COOPER NUCLEAR . STATION 10CFR50 APPENDIX E ON-SHIFT STAFFING ANALYSIS Revision 2 December 9, 2020 Prepared by: ~cn~Q!b., 12/9/2020

                    - ~Im Tanner          I      Date Peer Review  by:lt.1-Y':::: HlV(Ll,U..:v I 2/9/7o7D Tina Haynes Q   I      Date '
                   ~,,lr,u,I It               I;. 'I IU}~i)

COOPER Page 1

COOPER ON-SHIFT STAFFING ANALYIS REPORT TABLE OF CONTENTS I. INTRODUCTION .................................................................................................................................4 II. ANALYSIS

SUMMARY

.......................................................................................................................3 A. Emergency Plan Minimum Staffing .....................................................................................4 B. Other Commitments to Shift Staffing ....................................................................................5 C. Staffing Exceptions and Time Motion Studies (TMS) ........................................................ 5 D. Emergency Plan Tasks Not Analyzed ................................................................................. 6 Ill. ANALYSIS PROCESS .......................................................................................................................6 IV. ACCIDENT SCENARIOS ...................................................................................................................8 A. Accident Selection ...................................................................................................................8 B. Accident Scenarios included in the Analysis ..................................................................... 8 C. Accident Scenarios not included in the Analysis ............................................................... 9 V. GENERAL ASSUMPTIONS AND LIMITATIONS ......................................................................... 10 A. Notes and Assumptions applicable to Cooper OSA: .................................................... 10 B. NEI 10-05 Rev O Assumptions ............................................................................................ 10 VI. APPENDIX A - ANAL VZED EVENTS AND ACCIDENTS ........................................................... 13 VII. APPENDIX B - ON-SHIFT STAFFING ANALYSIS ................................................................... 15 A Accident Analysis #1 - Design Basis Threat (DBT) ....................................................... 15 B. Design Basis Accident Analysis #2 - Control Rod Drop Accident ............................... 21 C. Design Basis Accident Analysis #3 - Loss of Coolant Accident... ............................... 27 D. Design Basis Accident Analysis #5 - Fuel Handling Accident. .................................... 33 E. Design Basis Accident Analysis #6 - Main Steam Line Break ..................................... 39 F. Design Basis Accident Analysis #7 - Aircraft Probable Threat ..................................... 45 G. Design Basis Accident Analysis #8 - Control Room Evacuation and Shutdown ........ 51 H. Accident Analysis #9 - Station Blackout .......................................................................57 I. Accident Analysis #1 0 - LOCA/General Emergency with Release and PAR .............. 63 VIII. APPENDIX C- TIME MOTION STUDIES SUPPORTING THE STAFFING ANAL YSIS ....... 69 IX. OVERLAP OF TASKS ACTIVITIES OR OTHER CONFLICTS IDENTIFIED .......................... 73 A. Overlap Requiring Compensatory Measures .......................................... ..................73 REFERENCES ........................................... ........................................... .............................. 73 COOPER Page 2

COOPER ON-SHIFT STAFFING ANALYIS REPORT , ) XI. ORIGINAL STAFFING ANAL YIS TEAM ........................................................................................73 COOPER Page 3

COOPER ON-SHIFT STAFFING ANALYIS REPORT ,) INTRODUCTION This document is a revision to the Cooper Nuclear Station On-shrft Staffing Analysis (OSA) report dated July 20, 2015. This revision adds a clarifying not into the report identifying the process used for revising this document and Identifying it as part of the CNS Emergency Plan This revIsIon continues to satisfy the requirements of 10 CFR 50 Appendix E Section IV.A.9, which states that nuclear power licensees shall perform "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 the emergency plan." A structured approach was utilized to perform this analysis using the guidance found in NEI 10-05, Rev. 0, Assessment of On-Shift Emergency Response Organization Staffing and Capabilities. This analysis examined the capability of the minimum staff listed in Figure 5.2-1 of the Cooper Emergency Plan (E-Plan) to perform the actions for the key functional areas of events described In NSIR/DPR-ISG-01, Interim Staff Guidance - Emergency Planning for Nuclear Power Plants, until augmenting Emergency Response Organization (ERO) staff arrives In accordance with the E-Plan. NOTE: This report is considered part of the CNS Emergency Plan. Any changes made to this report must follow the guidance in CNS Administrative Procedure 0.29.1 "License Basis Document Changes". II. ANALYSIS

SUMMARY

The OSA team determined that an on-shift staff of fourteen (14) is required to respond to the most limiting accident scenario reviewed, main control room fire and shutdown at the remote shutdown panel. The on-shrft staff consists of individuals necessary to support each of the E-Plan functional areas or tasks*

Emergency Direction and Control
Plant Operations and Safe Shutdown (SSD)
Fire Fighting (FB)
Accident Assessment
Rad1at1on Protection and Chemistry
Notrf1cat1on/Commurncation
Technical Support
Access Control and Accountab1l1ty NEI 10-05 states It Is acceptable for certain functions or tasks to be assigned to personnel already assigned other functions/tasks These include Repair and Corrective Action, Rescue Operations and First Aid COOPER Page 4

COOPER ON-SHIFT STAFFING ANALYIS REPORT COOPER Page 5

COOPER ON-SHIFT STAFFING ANALYIS REPORT A. Emergency Plan Minimum Staffing Per 10 CFR 50.54 (q)(1 )(111), Emergency planning function means a capability or resource necessary to prepare for and respond to a radiological emergency, as set forth in the elements of section IV of Appendix E and, for nuclear power reactor licensees, the planning standards of§ 50.47(b). The following table indicates the result of the NEI 10-05 staffing analysis of on-shift personnel to perform the required emergency planning functions and the licensing basis requirement for each on-shrft position On-Shift Licensing Basis Staffing Position E-Plan Functional Area Requirement Analysis Results 50.54m Emergency Direction and Shift Manager (SM) E-plan 5.1 1, 5 1.2, 1 Control Figure 5 2-1 50.54m Control Room Supervisor (CRS) E-plan 5.1.1, 5.1 .2, SSD 1 Figure 5.2-1 E-plan 5.1.1, 5.1.2, Shrft Technical Engineer (STE) Technical Support 1 Figure 5.2-1 50.54m Reactor Operator #1 E-plan 5.1.1, 5.1 2, SSD 1 Figure 5 2-1 50.54m Reactor Operator #2 E-plan 5.1.1, 5.1.2, SSD 1 Figure 5.2-1 E-plan 5.1.1, 5.1 .2, Reactor Operator #3 FBL 1 Figure 5.2-1 E-plan 5.1.1, 5.1.2, Figure SSD Station Operator #1 1 5 2-1 E-plan 5.1.1, 5 1 2, Figure Station Operator #2 FB 1 5.2-1 E-plan 5.1 1, 5.1 2, Figure Station Operator #3 FB 1 5.2-1 E-plan 5 1 1, 5.1 2, Figure Notif1cat1ons and Communicator1 1 5.2-1 Communications E-plan 5.1.1, 5.1.2, Figure Dose Assessor 1 Dose Assessment 1 5 2-1 E-plan 5.1.1, 5 1.2, Figure Chem1stry/Rad1ation Chemistry/RP Technician 1 5 2-1 Protection E-plan 5.1.1, 5 1 2, Figure Utility Worker Fire Brigade #1 FB 1 5.2-1 E-plan 5.1.1, 5.1 2, Figure Ut1l1ty Worker Fire Brigade #2 FB 1 5 2-1 Per Security Access Control and Security Secunty Contingency Plan Contingency Accountability Plan TOTAL 14 1 Filled by staff qualified per TPP-101, Training Program Procedure and the Emergency Plan COOPER Page 6

COOPER ON-SHIFT STAFFING ANALYIS REPORT B. Other Commitments to Shift Staffing None. C. Staffing Exceptions and Time Motion Studies (TMS) 1 The primary responsibility for the Chemistry Technician is chemistry/radiochemistry sampling; however no chemistry job tasks were noted as being time critical or required during the first 60 minutes of any of the analyzed events. It is therefore acceptable to assign the Chemistry Technician the E-Plan function of dose assessment. No further analysis or TMS is required.

2. The Shift Manager 1s assigned the responsibility to make some event notifications such as the Duty Plant Manager, and Operations Manager. These notifications by phone are considered communications that are approximately one minute in length and are tasks for the Shift Manager.

No further analysis or TMS is required.

3. The analysis included a review of the implementation of the requirement to maintain continuous communications with the notification source during an aircraft threat in accordance with 10CFR50 54(hh) and Reg. Guide 1 214. There are no specific qualifications required to perform this task and the function 1s not required to be assigned in advance. The review of the Aircraft Threat Accident identrf1ed that there are sufficient personnel on-shift to perform this action during the event. Specifically, a licensed or non-licensed nuclear plant operator and a chemistry technician were all available to fill this function. No further analysis or TMS is required.

4 A Time Motion Study was completed to determine if the Shift Manager performance of ERO notif1cat1on would impact the capability to maintain emergency direction and control as Emergency Director. The TMS demonstrated that the Shift Manager was able to maintain oversight for emergency direction and control during the approximate 90 seconds that 1t took to activate the ERO using Dialog1cs ERO notification was evaluated as being an acceptable task for the Shift Manager. Details of the TMS are found in section VIII Appendix C COOPER Page 7

COOPER ON-SHIFT STAFFING ANALYIS REPORT D. Emergency Plan Tasks Not Analyzed

1. Repair and Corrective Action -. Repair and corrective action is defined as.

An action that can be performed promptly to restore a non-functional component to functional status (e g., resetting a breaker), or to place a component in a desired configuration (e.g.,

open a valve), and which does not require work planning or implementation of lockout/tagout controls to complete. In accordance with NE! 10-05 section 2.5, the analysis included a review of the repair and corrective action tasks. For the purpose of this analysis, the tasks were considered to fall into two broad categories:

Unplanned/unexpected actions that address equipment failures. These actions are contingent in nature and cannot be specified in advance
Planned/expected actions performed in support of operating procedure implementation, including severe accident management guidelines At CNS, Station Operators on shift will pnmarily be available to the Shift Manager to perform actions to assist in controlling/m1t1gating the event. Initial corrective actions are contained in and controlled by plant operating procedures The Station Operators have been trained to perform specrf1c actions within the Abnormal and Emergency Operating Procedures and maintenance support Is not required. Maintenance Is not included in the on-shrft staff. Repair and Corrective Action Is an acceptable collateral duty per the guidance of NEI 10-05 and was not analyzed

2. Rescue Operations and First Aid. In accordance with NEI 10-05 section 2.6, the analysis also included a review of rescue operations and first aid response. Rescue operations and first aid may be performed by shrft personnel assigned other functions At CNS the Search and Rescue function is handled by trained emergency response personnel. If station personnel are unaccounted for in the initial or subsequent personnel accountability, an emergency team will be assigned to locate and rf necessary, rescue them. Rescue operations and first aid response are acceptable collateral duties per the guidance of NEI 10-05 and was not analyzed Ill. ANALYSIS PROCESS The initial staffing analysis (Revision 0) was conducted by a Joint team of corporate Emergency Preparedness (EP) personnel and station personnel from the Operations, Training, Security, Chemistry and Emergency Preparedness (EP) departments. The team members are identified in Section XII of this report The emergency response to each event scenario was determined by conducting a tabletop of the event using the emergency plan and procedures and the applicable department procedures such as Operations emergency and abnormal operating procedures.

Each scenario was reviewed by the cross disc1pl1nary team to determine what plant actions and emergency plan implementation actions were required based on plant procedures prior to staff augmentation. These actions were then compared to the minimum staffing for Emergency Plan 1mplementat1on as described in the Emergency Plan Section 5 1 and Figure 5 2.1, ensuring that no actions were assigned to staff members that conflicted with either their dedicated emergency plan role or their dedicated operational role as appropriate. In cases where multiple __j tasks were assigned to an 1nd1v1dual In their role, the team evaluated tImIng of the tasks to ensure that they could be performed by the 1nd1v1dual in series within any spec1f1ed time requirements. COOPER Page 8

COOPER ON-SHIFT STAFFING ANALYIS REPORT The results of the analysis for each _of the scenarios are included in Section VIII, APPENDIX B- ON-SHIFT STAFFING ANALYSIS. Note that NSIR DPR-ISG--01 states that only OBA accidents "which would result in an emergency declaration" should be evaluated in the staffing assessment. Each of CNS DBA's were evaluated and classified according to ,ts FSAR Chapter XIV description. If the accident description alone did not result in a classification, the projected accident exclusion area boundary (EAB) dose found in the FSAR was utilized to determine if an EAL threshold would be exceeded within the first 60 minutes using the Abnormal Rad Level EAL thresholds. In cases where several projected dose rates were provided or release data was not detailed significantly to determine an EAL, the assessment used the radiological consequences associated with the realistic case in accordance with NEI 10-05. COOPER Page 9

COOPER ON-SHIFT STAFFING ANALYIS REPORT IV. ACCIDENT SCENARIOS A Accident Selection

1. The OSA scenarios were chosen using the guidance of NEI 10-05 and NSIR/DPR-ISG-01, "Interim Staff Guidance - Emergency Planning for Nuclear Power Plants." The evaluation considered the station Design Basis Accidents (OBA) described m the FSAR along with additional scenarios specified by the guidance documents. The scenarios considered were.

Design Basis Threat (DBT) ground assault
OBA Control Rod Drop Accident
OBA Loss of Coolant Accident
Accidents That Result in Direct Release to Secondary Containment
OBA Fuel Handling Accident
OBA Main Steam Line Break Accident
Aircraft Probable Threat
Fire Requ1ring Evacuation of the Control Room and Plant Shutdown From Alternate Location
Station Blackout, (SBO)
LOCNGeneral Emergency with release and PAR
LOCA with entry into Severe Accident Management
Fire Resulting in Reactor Trip (NFPA 805 fire)

B. Accident Scenarios included m the Analysis

1. Design Basis Threat (DBT) as described in NEI 10-05

Land and/or waterborne Hostile Action directed against the Protected Area by a Hostile Force This event assumes the threat is neutralized 1mmed1ately when inside the protected area fence, no s1gnrf1cant damage to equipment or systems that require corrective actions before the ERO is staffed, no radiological release, and no fire that requires f1ref1ghting response before the ERO Is staffed EAL is based on the event.

2 Control Rod Drop Accident as described m FSAR XIV-6 Section 6.2

One control rod drops and reactor trips on APRM 120% flux scram Credit is not taken for MSIV closure. Release pathway Is through condenser, turbine building, then to environment. EAL is based on condenser release EAB dose information m FSAR.

3. Loss of Coolant Accident as described m FSAR XIV-6 Section 6.3

Double ended guillotine break of a rec1rculat1on line. All ECCS operate as designed. EAL Is based on fission product barrier EALs.

4 Fuel Handling Accident as described In FSAR XIV-6 Section 6 4

The accident involves a dropped fuel bundle on top of the core. Initial EAL Is based on event COOPER Page 10

COOPER ON-SHIFT STAFFING ANALYIS REPORT 5 Main Steam Line Break Accident as described In FSAR XIV-6 Section 6.5

Break of one main steam line and followed by MSIV closure. Release Is to the atmosphere through the top of the turbine building (puff release). Reactor coolant activity at TS limit.

No EAL on the accident but on the USAR EAB 2-hr dose.

6. Aircraft Probable Threat (50.54hh)

Notrf1cat1on is received from the NRC that a probable aircraft threat exists (>5 minutes,

                                 <30 minutes). EAL is based on the event.

7. Control Room Fire and Remote Shutdown.

A fire occurs in the main control room requiring the evacuation and the procedure implemented to perform shutdown from the alternate shutdown panels. EAL is based on the event.

8. Station Blackout.

A loss of all offsite AC power occurs and the failure of the emergency diesel generators.

EAL is based on the event.

9. General Emergency with release and PAR

Assume GE condition with dose projection 1nd1cat1on of a release greater than the protective action guideline (PAG)

C Accident Scenarios not included in the Analysis

1. Accidents That Result in Direct Release to Secondary Containment

FSAR includes 2 varieties of accidents: (1) Failures of the reactor coolant pressure boundary inside Secondary Containment and (2) failure involving fuel that is located outside the Primary Containment but inside the Secondary Containment. These accidents are not described since they are less severe than similar accidents in other categories No Analysis is required.

2 Implement SAMG

NEI 10-05 Section 2.11 states that the analysis of the ability to implement SAMG focuses on the reasonably expected initial mitigation action that would be performed by on-shift personnel other than licensed and non-licensed operators. A review of the SAMGs associated with the initial site-spec1f1c Candidate High Level Actions concluded that no actions would require on-shrft personnel other than licensed and non-licensed operators.

No analysis required. 3 NFPA 805 Fire

Fire leading to reactor trip with complications. The control room fire and alternate shutdown requires more resources and actions outside the control room than this fire. The team concluded the control room fire bounds the fire analysis. No add1t1onal analysis required.

COOPER Page 11

COOPER ON-SHIFT STAFFING ANALYIS REPORT GENERAL ASSUMPTIONS AND LIMITATIONS A. Notes and Assumptions applicable to Cooper OSA: 1 The RP and Chemistry tasks reviewed were those directed by the Shift Manager to support actions in Abnormal Operating Procedures (AOP), Emergency Operating Procedures (EOP), and Emergency Plan Implementing Procedures (EPIP). Any additional tasks directed by the Technical Support Center (TSC), Operations Support Center (OSC), or Emergency Operations Facility (EOF) procedures were not reviewed.

2. The OSA team determined RP/Chem Tech tasks performance is directed and prioritized by the Shrft Manager The time the RP/Chem Tech is directed to perform a task and the amount of time taken to complete tasks are estimated. No Chemistry samples are required by Tech Specs within the 60 minute penod after the declaration. Since the Shift Manager directs when the tasks are performed, there are no overlapping RP or chemistry tasks.

3. All crews have one individual filling the SM and one 1nd1v1dual filling the STE roles therefore the analysis did not consider using a dual-role individual

4. While the augmentation time used for this analysis was based on the time 1dentrf1ed in the station's emergency plan (60 minutes), no additional tasks requiring the augmented ERO were noted as being required of the on-shift ERO for the first 90 minutes following the emergency declaration B NEI 10-05 Rev 0 Assumptions

1. Response time used for this analysis was the maximum acceptable number of minutes elapsed between emergency declaration and the augmented ERO position holder at a location necessary to relieve an on-shift position of the emergency response task. (60 min.).

2 On-shrft personnel complement was l1m1ted to the minimum required number and composItIon as described in the site emergency plan. If the plan commitments allow for different minimum staffing levels (e g., a variance between a normal dayshift and a backshrft), use the staffing with the smallest number of personnel. 3 Although the temporary absence of a posItIon may be allowed by Tech Specs, the analysis was performed assuming that all required on-shrft positions are filled.

4. Event occurred during off-normal work hours where ERO was offsite and all required minimum on-shrft posItIons were filled.

5. On-shrft personnel reported to their assigned response locations within timeframes sufficient to allow for performance of assigned actions

6. On-shrft staff had necessary Rad1at1on Worker qualif1cat1on to obtain normal dosimetry and enter the rad1olog1cal control area (RCA) (but not locked high or very high radiation areas) without the aid of a RP technician COOPER Page 12

COOPER ON-SHIFT STAFFING ANALYIS REPORT ~ 7. Personnel assigned plant operations and safe shutdown meet the requirements and guidance (analyzed through other programs such as operator training) and was evaluated as part of this assessment unless a role/function/task from another major response area was assigned as a collateral duty.

8. In-plant (manual) safety related operator actions to manipulate components and equipment from locations outside the control room to achieve and maintain safe shutdown was done by a member of the on-shrft staff as defined in the unit's Tech Specs.

9. Fire brigade (FB) staff performance is analyzed through other station programs (e.g., fire drills) and was not evaluated as part of this assessment unless a role/function/task from another maJor response area was assigned as a collateral duty.

10. Individuals holding the position of RP techn1c1an or Chemistry technician are qualrfied to perform the range of tasks expected of their position.

11. Security was not evaluated unless a role or function from another major response area was assigned as a collateral duty.

12. Communications, briefings, and peer checks are acceptable collateral duties 13 All on-shift staff positions were evaluated, even if they had no known collateral duties, to ensure they can perform the tasks assigned to them. [Ref NSIR/DPR-ISG-01]

14 . The Staffing Analysis specified the resources available to perform "Repair and Corrective .~) Actions" and "Rescue Operations and First Aid" but these may be assigned as collateral duty to a designated on-shrft responder

15. For assessment purposes, NRC notifications were treated as a continuous action per 10CFR50.72(c)(3) and 73.71(b)(1). This means once the 1nit1al NRC communications are established, the NRC will request an open line be maintained with the NRC Operations Center.

16 OBA (postulated accident, Condition IV event, or limiting fault) Is considered as "Unanticipated occurrences that are postulated for accident analysis purposes but not expected to occur during the lrfe of the plant. A postulated accident could result in sufficient damage to preclude resumption of plant operation. As a result, a greater number and variety of actions would need to be implemented by plant personnel." 17 Unless otherwise specrf1ed in NSIR/DPR-ISG-01, Interim Staff Guidance - Emergency Planning for Nuclear Power Plants, or by the FSAR initial cond1t1ons of a OBA analysis, 1t was assumed that the unit was In Mode 1, Power

18. DBT assumed a hostile force breached the protected area fence but was neutralized with no adverse consequences to plant safety Damage inflicted on plant systems, structures and components was not sufficient to prevent safe shutdown or cause a radiological release. There was no fire s1gnrf1cant enough to warrant f1ref1ghting efforts prior to amval of offsIte resources and/or the augmented ERO 19 The Staffing Analysis used OBA analysis assumptions, inputs, timing of events, plant protective response, and specrfied manual operator actions and their tImIng, as documented In the FSAR.

COOPER Page 13

COOPER ON-SHIFT STAFFING ANALYIS REPORT ~I 20. In cases where a OBA analysis included a radiological release, and the starting point of the release was not clearly defined, the staffing analysis assumed that the release began 15-minutes after the initiating event.

21. Severe Accident Management Guideline (SAMG) - It is suff1c1ent to simply assume that the accident progressed to conditions requiring a severe accident response; it is not necessary to determine specific failures and the accident sequence.

22. SAMG - The actions analyzed included those that implement the initial site-specrfic actions assuming the core is not ex-vessel (i.e., no reactor vessel failure), and there is no actual or imminent challenge to containment integrity.

COOPER Page 14

COOPER ON-SHIFT STAFFING ANALYIS REPORT

 .r- I. APPENDIX A - ANAL VZED EVENTS AND ACCIDENTS 1  DBT                Land and/or        1      NEI 10-05             Site Area         Yes waterborne                                      Emergency HOSTILE                   ISGIVC ACTION directed against the Protected Area by a HOSTILE FORCE.

2 DBA Control Rod Drop 1 FSAR XIV-6 Secbon 6.2 Site Area Yes Accident Emergency 3 OBA Loss of Coolant 1 FSAR XIV-6 Section 6.3 Site Area Yes Accident Emergency 4 OBA Accidents that 1 Not described m FSAR N/A 2 No result in radioactive matenal release directly to the Secondary Containment 5 OBA Fuel Handling 5 FSAR XIV-6 Section 6 4 Alert Yes Accident j6 OBA Mam Steam Line FSAR XIV-6 Section 6.5 Unusual Event Yes Break 7 Assumed for Aircraft Probable 10CFR50 54hh (1) Alert Yes analysis purpose Threat RG 1.214 8 Assumed for Control Room FSAR 14.6 1 Alert Yes Analysis Purpose Evacuation and Alternate Shutdown (fire m mam control 10CFR50 NFPA 805 Fire room) Hazard Analysis 9 Assumed for Station Blackout 1 ISGIVC Site Area Yes analysis purpose Emergency 10 Assumed for LOCA- General ISGIV.C General Emergency Yes Analysis Purpose Emergency with radiological release and PAR 11 Assumed for LOCA with entry ISG IVC General Emergency No3 Analysis Purpose into severe accident procedures 12 Assumed for NFPA 805 Fire ISG IVC Site Area No4 Analysis Purpose with Reactor Trip Emergency

~ )

1 Plant mode per FSAR or assumed for analysis purpose COOPER Page 15

COOPER ON-SHIFT STAFFING ANALYIS REPORT ,,......-...'{lcludes 2 vanet,es of accidents: (1) Failures of the reactor coolant pressure boundary inside Secondary ontainment and (2) failure involving fuel that Is located outside the Primary Containment but inside the Secondary Containment. These accidents are not described in the FSAR since they are less severe than similar accidents in other categories. Analysis is not required. 3 CNS does not meet the NEI 10-05 intent for the analysis of implementing SAMG. NEI 10-05 Section 2.11 states that the analysis of the ability to implement SAMG focuses on the reasonably expected initial mitigation action that would be performed by on-shift personnel other than licensed and non-licensed operators. The actions assessed by NEI 10-05 are those which implement the initial site-specific Candidate High Level Action assuming the core is not ex-vessel (1 e., no reactor vessel failure), and there is no actual or imminent challenge to containment integrity. CNS does not include maintenance qualified to perform maintenance job tasks in minimum staffing and any response actions would be performed by operators. 4 NFPA 805 Fire Is bound by the Control Room Fire and Alternate Shutdown. Analysis Is not required. COOPER Page 16

COOPER ON-SHIFT STAFFING ANALYIS REPORT APPENDIX 8 - ON-SHIFT STAFFING ANALYSIS A. Accident Analysis #1 - Design Basis Threat (DBT)

1. Accident Summary

Land and/or waterborne HOSTILE ACTION directed against the Protected Area by a HOSTILE FORCE. Assume adversary characteristics defined by the Design Basis Threat. Security Code Red condition

2. Accident Specific Assumptions Made

This event assumes the threat is neutralized immediately when 1ns1de the protected area fence, no significant damage to equipment or systems that reqwre corrective actions before the ERO is staffed, no radiological release, and no fire that requires firefighting response before the ERO is staffed.
Assume at power in Mode 1
Security notifies the Shift Manager of condition Security Code RED.
Assume all non-security staff is located inside the protected area at their normal work station when the event occurs.
Assume all systems function and the core remains covered. No fuel damage and no release.

3. Procedures for Accident Response

5.5SECURITY
5.7.1, Emergency Classification
5.7.2, Emergency Director EPIP
5.7.6, Notification COOPER Page 17

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing.

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                 !Shift Manager                                                 60                           T5/l6              No                  Yes Plan i::igure 5.2-1                                                         T5/L7 T5/L 11 T5/L 14 Emergency Control Room 2
                 !Supervisor Plan                                     NIA                          T2/L2               No                   No Figure 5 2-1 Emergency Shift Technical                                                                           T2/l3 3                         Plan                                      60                                              No                   No Engineer (STE)                                                                             T5/L 10 Figure 5.2-1 Emergency Reactor 4

Operator #1 Plan N/A T2/L4 No No

                                      ~iqure 5.2-1 Emergency Reactor 5                         Plan                                     NIA                          T2/L5               No                   No

) Operator#2 Figure 5.2-1 Emergency Reactor 6

Operator#3 Plan NIA NIA No No Figure 5.2-1 Emergency Station Operator

                 #1 Plan                                      N/A                          T2/l6               No                   No figure 5.2-1 Emergency
                !Station Operator 8
                 #2 Plan                                      NIA                            N/A               No                  No
                                     ~Igure 5.2-1 Emergency
                !Station Operator*                                                                           N/A 9                         Plan                                       60                                              No                  No W3 i::1gure 5.2-1 Emergency                                                              T5/l8 10   !Communicator Plan                                               60                          T5/L9               No                  No IF1qure 5.2-1                                                          T5/L 13 Emergency RP/Chem Tech 11                         Plan                                       60                           NIA                No                  No 1#1 Figure 5.2-1 Emergency 12    Dose Assessor Plan                                              60                           NIA                No                  No Figure 5.2-1 Utility Fire         Emergency 13    Brigade              Plan                                      N/A                           NIA                No                  No Member #1           Figure 5.2-1 Utility Fire        Emergency 14    Bngade              Plan                                       NIA                           N/A

,___,,) 1Member#2 Figure 5 2-1 No No COOPER Page 18

COOPER ON-SHIFT STAFFING ANALYIS REPORT ~ Security 15 ISecunty Contingency 60 T5/L 15 No No IPian COOPER Page 19

COOPER ON-SHIFT STAFFING ANALYIS REPORT 1 Licensed Operator Training Shift Manager Shift Manager Program 2 Licensed Operator Train mg Unit Supervisor Control Room Supervisor Program 3 Shift Technical Engineer Licensed Operator Training

                 !Shift Technical Advisor                                                     (STE)            Program 4                                                                                                   Licensed Operator Training
                 !Reactor Operator #1                                             Reactor Operator #1          Program 5                                                                                                   Licensed Operator Training Reactor Operator #2                                              Reactor Operator #2          Program Non-Licensed Operator 6    ~ux1hary Operator #1                                             Station Operator #1          Training Program 7    Auxiliary Operator #2                                                          NIA            NIA 8    Other needed for Safe Shutdown                                                 NIA            NIA 9    Other needed for Safe Shutdown                                                 NIA            NIA

'.~) 10 Other needed for Safe Shutdown NIA NIA Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs rf Applicable

         ;Line#
                         ..                               :                                                           Task Analysis

Generic ... Title/Ro.le

                                                             ... ,.,... ',,       '.. ,,_p~~Shift,P.o~ition :*
                                     ...                '                                          .*,.            G9!1trolllng*l\(ethocl_..

11 Mechanic NIA NIA 12 Electrician NIA NIA 13 l&C Technician NIA NIA 14 Other NIA NIA 15 Other NIA NIA COOPER Page 20

COOPER ON-SHIFT STAFFING ANALYIS REPORT

       *-                       -                                                            ..      ~  ..

CNS TABLE 3 - FIREFIGHTING ANALYSIS # 1 DBT Securiti Threat Line Performed by Task Analysis Controlling Method 1 N/A Fire Protection Program 2 N/A Fire Protection Program 3 N/A Fire Protection Program 4 N/A Fire Protection Program

5 N/A Fire Protection Program Note : This accident does not include the need for firefighting , first aid or search & rescue . COOPER Page 21

COOP ER ON-SHIFT STA I ING ANALYIS REPORT

CNS TABLE 4- RADIATION PROTECTION AND CHEMISTRY Analysis # 1 DBT Securitll Threat L Position Perform ing Performance Time Period After Emergency Declaration (minutes) I Function/ Task N 0-5 5-10 10-15 15-20 20-25 25-30 30-35 35-40 40-45 45-50 50-55 55-60 60-65 65- 70- 75- 80- 85-90 E 70 75 80 85 1 In-Plant Survey: N/A 2 On-site Survey: N/A 3 Personnel Monitoring: N/A 4 Job Coverage: N/A 5 Offsite Rad Asse ssment: N/A 6 Other site specific RP {describe) : N/A 7 Chemistry Function task

              #1 {describe)

N/A 8 Chemistry Function task

              ~2 (describe)

N/A Note: No chemistry or RP job functi on tasks fo r the conditions described in the DBT assum ptio ns. The RP/Chem Tech takes cover as directed. COO PER Page 22

COOPER ON-SHIFT STAFFING ANALYIS REPORT CNS TABLE 5 - EMERGENCY PLAN IMPLEMENTATION - Analysis # 1 DBT Securitll'. Threat Function / Task On-Shift rrask Analysis Controlling Method Linell Position Declare the emergency classification level Emergency Planning Training 1 Shift Manager (ECL) Program I EP Drills Approve Offsite Protective Action 2 Recommendations N/A N/A Emergency Planning Training 3 Approve content of State/local notifications Shift Manager Program 4 !Approve extension to allowable dose N/A N/A Licensed Operator Training Notification and direction to on-shift staff 5 Shift Manager Program I Emergency Planning (e.g., to assemble, evacuate, etc.) Training Program Emergency Planning Training 6 ERO notification Shift Manager Program Licensed Operator Training 7 ~bbreviated NRC notification for DBT event Shift Manager Program Emergency Planning Training 8 Complete State/local notification form Communicator Program Emergency Planning Training 9 Perform State/local notifications Communicator Program Licensed Operator 10 Complete NRC event notification form STE rrraining Program Emergency Plann ing Training 11 Activate EROS Shift Manager Program 12 Offsite radiological assessment N/A N/A Emergency Plann ing Training 13 Perform NRC notifications Communicator Program Perform other site-specific event Licensed Operator 14 notifications (e.g., Duty Plant Manager, Shift Manager INPO , ANI , etc.) rrraining Program Security Training Program / EP 15 Personnel Accountability Security Officer Drills COOPER Page 23

COOPER ON-SHIFT STAFFING ANALYIS REPORT B. Design Basis Accident Analysis #2 - Control Rod Drop Accident

1. Accident Summary

The control rod drop accident (CRDA) results from an assumed failure of the control rod-to-drive mechanism couphng after the control rod (very reactive rod) becomes stuck in its fully inserted position It Is assumed that the control rod dnve Is then fully withdrawn before the stuck rod falls out of the core. The control rod velocity l1m1ter, an engineered safeguard, limits the control rod drop veloCJty The resultant radioactive material release is maintained far below the guideline values of 10CFR50 67
Loss of offsite power coincidental with CRDA.
Radionuclides are released from damaged fuel rods to the main condenser Single release path Is modeled from the leakage of main condenser at 1% volume per day to the turbine building. Activity is distributed throughout the turbine building and passes directly to the environment (with no mIxmg or holdup in the TB volume) as a diffuse ground level release

2. Accident Specific Assumptions Made

EAL is based on FSAR EAB dose 1nformat1on.

3. Procedures for Accident Response

5.1 RAD, Building Radiation Trouble
2.4OG, Offgas Abnormal
5.2Fuel, Fuel Failure

-~)

5.3EMPWR, Emergency Power During Modes 1, 2, and 3
2.1 .5, Reactor Scram
5. 7 16, Release Rate Determination
5.7.17, Dose Assessment
5.7.19, On-site Rad1ological Monitoring
5.7.1, Emergency Classification
5.7.2, Emergency Director EPIP
5.7 6, Notification
5. 7.10, Personnel Assembly and Accountability
9.EN-RP-142, Failed Fuel Response COOPER Page 24

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing.

                                              *.. *CNS T:AB1:-;E j *_-;-,*QN;.Sl;:IIF:Jq~,qs,moNS. \' : . ,.  ; ' -              ,_
                                            *:*:Analysis # 2 Control

_, * > r - ;. - -- ~ ,- _-: Roa .Drop Accident

                                                                              *L J_     ~--:* *--* * *   '  '~
                                                                                                                     * .. ~ * >- r
       ..,.-,- ...l Lih~ #Kln~hift Posttion
       ;      a                        *,

T2/L1 T5IL 1 Emergency Plan T5IL3 1 !Shift Manager 60 No Yes Figure 5.2-1 T5IL5 T5IL6 T5IL 14 Control Room Emergency Plan 2 NIA T2/L2 No No Supervisor Figure 5.2-1 Shift Technical T2/L3 Emergency Plan 3 60 T5IL 10 No No Engineer (STE) Figure 5.2-1 T5IL 11 Emergency Plan 4 Reactor Operator #1 NIA T2/L4 No No Figure 5.2-1 Emergency Plan 5 Reactor Operator #2 NIA T2/L5 No No Fiqure 5.2-1 Emergency Plan 6 Reactor Operator #3 NIA NIA No No Fiqure 5.2-1 Emergency Plan 7 Station Operator #1 NIA T2/L6 No No Fiqure 5 2-1 Emergency Plan NIA NIA 8 Station Operator #2 No No Figure 5.2-1 Emergency Plan 60 NIA 9 Station Operator #3 No No Figure 5 2-1 T5/L8 Emergency Plan 10 Communicator 60 T5/L9 No No Figure 5.2-1 T5/13 T4/L 1 Emergency Plan 60 11 RP/Chem Tech #1 T4IL7 No No Figure 5 2-1 Emergency Plan T5IL 12 12 Dose Assessor 60 No No Figure 5 2-1 Utility Fire Brigade Emergency Plan 13 N/A NIA No No Member #1 Figure 5.2-1 Ut1l1ty Fire Brigade Emergency Plan 14 NIA NIA No No Member#2 Figure 5.2-1 Security 15 Security 60 T5IL 15 No No Continqency Plan COOPER Page 25

COOPER ON-SHIFT STAFFING ANALYIS REPORT

                       --    _, *: *, , CNS TA_Bt,:,~--~ _o:. p~~T O_PE~                nq~s       ~ 'S~Fl;:       ~t:f U"f99~- -.                ~         * -- .
                           , . ,. * .\ -. ,. ** ".;*- * *-* On~.i.,n~.-.O!1e*CQhtrol*Room *. .,, -. " *--,: - ..                                     -t       :,**   1
                                *,     .           . :_ ANALYSIS ,#-2 Control Rod D'i-op-Accident . .. , . . --,.                                       . .

- ) ' ": ~ *-~' - f} - ,_-_ - **.-,, ';L - ~ -,_ ,~.,*~*, ~ ,,- , ,- ,, . ::~,:;--- --;_.,:<:-;,:A:>. ,*,: . ' .... /': , y

l M_in!~.u~;<?813ryi_v~~~- G!"9W N~~~~§'Y. {ci l!!i~i~~e.~!_AOP~ -a_n'ff -~Q~s 6:~:§1M<~f,if_ ~PP,l!:c:a~j_~ -

Line# !Generic Title/Role On-Shift Position rrask Analysis

                                                                                                                                    !Controlling Method licensed Operator 1         !Shift Manager                                               Shift Manager rTraining Program Licensed Operator 2         Unit Supervisor                                               Control Room Supervisor rrra1nmg Program Shift Technical Engineer                          Licensed Operator 3         Shrft Technical Advisor (STE)                                            rrraining Program licensed Operator 4         Reactor Operator #1                                          Reactor Operator #1 rrra1rnng Program licensed Operator 5         Reactor Operator #2                                          Reactor Operator #2 rTraining Program Non-Licensed Operator 6        1Auxil1ary Operator #1                                       !Station Operator # I Training Program 7        1Aux1hary Operator #2                                        IN/A                                               N/A

_--) 8 !Other needed for Safe Shutdown NIA N/A 9 Other needed for Safe Shutdown NIA NIA 10 !Other needed for Safe Shutdown IN/A IN/A Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs rf Applicable Line# Generic Tltle/Rol_e . * : - - ' - On-Shift. Position ' : rTask Analysis

       .     - ..              -                     -*            J" ,_ -
                                                                                                                                     ~~trolling-M~hQd:

_

11 Mechanic NIA N/A 12 Electrician N/A N/A 13 l&C Technician N/A N/A 14 Other N/A N/A 15 Other N/A N/A COOPER Page 26

COOPER ON-SHIFT STAFFING ANALYIS REPORT ') Fire Brigade

                                 .:     *. : - CNS' TABLE:-3:.... FIREFIGHTING*-,,,*;.- *- .. *: *
                             .'. j,. _- ANAhYSIS .tl-2.:controLR.od Dto~~rdent .: _, _- .. ; : ._*. '< ' -, -_~; .. '-;.

NIA NIA 2 NIA NIA 3 NIA NIA 4 'NIA NIA 5 NIA NIA The CRDA does not include the need for firefighting.

COOPER Page 27

COOPER ON-SHIFT STA~IING ANALYIS REPORT J

                                                                                                                                                  . ' . ~;,_;:

C,NS' TABlE 4 - RADIATION'.PROTECTION,ANO 'CHEMISTRY~::**\,, -

                                                                                                                                                                                     ,f~,.,.        'J- ......  "'
                                                                                                                                                                                                                       "~
                                               -             Analysis*f2 Cohfrol.Rnrl, Drnn Accident * ,, . ,*", _. * :-* :
                                                                                                                                        ..      '-* - ~       .....                       - -~  "            ~

L Position Performing ' 0

                                                                                                                                                       ',             '                   -~ :
                                                            .,    Perfqrmaric;e Ti~~ -~eJJod..Aft~r:.~111~*rgeniy Deplaratl9n, (mi~ute~t                                                             *              -"

J ' function /Task **. *' '

                                                                                                                                                                 ..,     ' ~'   ,   ' ' ) . ., >J",                 ,,

N I

                                                                                               ' ~   ' (
                                                                                                                                 -*~5'-" '7(j.:.\ I.. 7.5-                      '

80-, 1

      -    -~-          -

. 0-5 5-10 10-15 15-20 2!)-25 25-30 30-35 3&-40 4'0-45 [45-50 50:-55 155-60 l60:.as

                                                  ' ' *'.                                        ' ' -      ' .  -             :  _70     _.:     t,5_:       .80
                                                                                                                                                                                    .. 85'_*                  85-90 1     I In-Plant Survey.

Applicable steps of EN-X X X X X X X X X RP-142, Failed Fuel

Response

2 On-site Survey. N/A 3 Personnel Monitoring: N/A 4 Job Coverage' N/A 5 Offsite Rad Assessment

                   'Included in Table 5) 6     Other site specific RP (describe) N/A 7     Chemistry Function task
                  #1 (describe) - N/A 8     Chemistry Function task
                  #2 (describe)- N/A
                  *Performance Times are estimated Rx coolant sample as directed by the SM Is not required until 2-6 hours after shutdown.

Procedure directs Offgas sample (not time cntical) but it isolates. Sample not required during the 60 minutes COOPER Page 28

COOPER ON-SHIFT STAFFING ANALYIS REPORT

                                .. CNS.TABLE 5.::. EMERGENCY PLAN" IMPLEMENTATION *: -                                                          ..  ., ;- -   ~ -. /.- ~-      '
              .,                -~ * .: .. Anaiy&is' # 2*control'RbcHlrop Accident, **, * *_,* * '* ~- -~
                                       <      I~    J' ,   > * .. - ,  '..,* **   *  -*- *   *  '/ -      *    *, *       *    ' '-   ,,. ,
                                                                                                                                               * '* .I;

_ !\ .~* /,r..~-~~

                                                                                                                                                                      '*   r
                                          ~ -    - ,,                           **  ---    -   c__, . ;

_* -~:_' 'h.. :,...,~:..,. ~**'

       'l;Jr(e
         ~#.

Emergency Planning Declare the emergency classrficat1on level 1 Shift Manager Training Program/ EP

                 'ECL)

Drills

                 !Approve Offsite Protective Action 2     Recommendations NIA                                                   NIA Emergency Planning 3     !Approve content of State/local notrf1cat1ons        Shift Manager Training Program 4     !Approve extension to allowable dose                  NIA                                                  IN/A Licensed Operator Notification and direction to on-shift staff                                                               [Training Program /

5 IShlft Manager (e.g., to assemble, evacuate, etc.) Emergency Planning [Training Program Emergency Planning 6 ERO notif1cat1on !Shift Manager rrraming Program 7 Abbreviated NRC notrf1cation for DBT event NIA N/A Emergency Planning 8 Complete State/local notification form !Communicator rTra1rnng Program Emergency Planning 9 Perform State/local notrf1cations !Communicator [Training Program ~) 10 Complete NRC event notrf1cation form ISTE Non-Licensed Operator rTraining Program Emergency Planning 11 Activate EROS ISTE [frammg Program Emergency Planning 12 Offs1te rad1olog1cal assessment !Dose Assessor rTraining Program Emergency Planning 13 Perform NRC notrf1cat1ons Communicator rT"rammg Program Perform other site-specrf1c event Licensed Operator 14 notJf1cat1ons (e.g, Duty Plant Manager, !Shift Manager rT"rainmg Program INPO, ANI, etc)

                                                                                                                            !Security Training Program 15     Personnel Accountability                             ISecunty V EP Drills COOPER                                                                                                                                                                           Page 29

COOPER ON-SHIFT STAFFING ANALYIS REPORT C. Design Basis Accident Analysis #3 - Loss of Coolant Accident

1. Accident Summary

This event involves the postulation of a spectrum of pIpIng breaks inside containment varying in size, type, and location. Break of a large recirculation pipe represents the limiting pipe break inside the containment Reactor scram, MSIVs isolate, ECCS initiates and injects.
Worst case 2-hour period for exclusion area boundary dose occurs from 1.3 to 3.3 hours of 1.002 rem TEDE

2. Accident Specific Assumptions Made

Loss of offsite power at the initiating event and is not restored during event.
No operator action for the first 10 minutes. Post-LOCA manual valve manipulations are after the 1rnt1ating alarm Is received.
Classification is made on fission product bamer EAL.

3. Procedures for Accident Response

5.2FUEL, Fuel Failure
5.1 RAD, Building Radiation Trouble
2.1.5, Reactor Scram
5.7.1, Emergency Classification
5 7 2, Emergency Director EPIP
5.7.6, Notification
5.7.10, Personnel Assembly and Accountability
5 7 14, Stable Iodine Thyroid Blocking (Kl)
5.7 16, Release Rate Determmabon
5.7.17, Dose Assessment
5 7.19, On-site Radiological Morntonng COOPER Page 30

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains staffing identified in Section II.A, OSA Minimum Staffing

          .-.- - - . -:. >- ,-~ . *_ ~-. :- - *CNST.ABLE*1*..:,.oN:s:HIFTPOSr@ONS *.,_.-.r . _:*.--.                                                        .. * ... *., *  ...:*
             -        * -*  '     I,
                                            ~
                                                           *,Ancilysis.ti;~-,L~~ ofGb'oiarit ~ccideiit. :* :_ *~-- * ~- ;
                                                                        ,. * * - - .-- ~ * . *~ - * * "., ,,- .- ~ *-  ~    - ,-_*..,. -.. . -

* ,- *.., * * ;.~ " * '

      *Un~ b~~h-ift :: .:_ .- .~- ~~is:_.'.: ..*_: *\
      * #   :Posit~6n.:-. :' ;; Docinnent* .
                         ; ..,_,'   :'*; - -- , = - ...... _,. : ,_

T2/L 1 T3/L 1 T5/L 1 Emergency T5/L3 1 Shift Manager Plan 60 No Yes T5/L5 Figure 5.2-1 T5/L6 T5/L 14 T5/L 11 Emergency Control Room 2 Plan N/A T2/L2 No No

             !Supervisor Figure 5.2-1 Emergency Shift Technical                                                                                        T2/L3 3                                         Plan                              60                                                         No                       No Engineer (STE)                                                                                        T5/L 10 F1qure 5.2-1 Emergency Reactor 4

Operator#1 Plan N/A T2/L4 No No Figure 5.2-1 Emergency

             !Reactor 5

Pperator#2 Plan N/A T2/L5 No No Figure 5.2-1 Reactor Emergency 6 Plan N/A N/A No No Operator # 3 Figure 5.2-1 Station Operator EmePrlgency 7 #1 an N/A T2/L6 No No Fiqure 5.2-1 Emergency 8 S tatIon Operator Plan N/A N/A No No

             #2                          Figure 5.2-1 Emergency 9    Station Operator                    Plan                               60                                  N/A No                       No
             #3                          F1qure 5.2-1 Emergency                                                                 T5/L8 10 Communicator                           Plan                               60                                T5/L9                    No                       No Figure 5.2-1                                                                 T5/13 T4/L 1 Emergency RP/Chem Tech                                                           60                                T4/L4 11                                        Plan                                                                                          No                       No
             #1                                                                                                       T4/L6 Figure 5.2-1 Emergency 60                              T5/L 12 12 Dose Assessor                          Plan                                                                                          No                       No Figure 5 2-1 Utility Fire                   Emergency 13 [Brigade                               Plan                             NIA                                  N/A                     No                       No Member #1                   Figure 5.2-1 Uti11tyF1re                    Emergency                                                                   NIA 14                                                                         N/A                                                          No                       No Brigade                             Plan COOPER                                                                                                                                                                               Page 31

COOPER ON-SHIFT STAFFING ANALYIS REPORT Member#2 Figure 5 2-1 Security 15 Security Contingency 60 T5/L15 No No Plan COOPER Page 32

COOPER ON-SHIFT STAFFING ANALYIS REPORT Line# Generic Title/Role !On-Shift Position tTask Analysis lcontrolllng Method

                                                                                      !Licensed Operator 1    Shrft Manager                                 Shift Manager trraining Program Licensed Operator 2    Unit Supervisor                        Control Room Supervisor Training Program Shift Technical Engineer IL1censed Operator 3   !Shift Technical Advisor (STE)             tTraining Program
                                                                                      !Licensed Operator 4    Reactor Operator #1                        Reactor Operator #1 tTrainmg Program Licensed Operator 5    Reactor Operator #2                        Reactor Operator #2 trraining Program 1Non-L1censed Operator 6   1Aux1l1ary Operator #1                      Station Operator #1 tTraming Program 7   Auxiliary Operator #2                                N/A              N/A

,J 8 9

                 !Other needed for Safe Shutdown Other needed for Safe Shutdown N/A N/A N/A N/A 10    !Other needed for Safe Shutdown                      N/A              N/A Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs rf Applicable Line*# IGeneric'Xitle/Role \
                        '*

L '

                                                         !Ori-Shift Position
                                                                                    ,'trask
                                                                                         ,   Analysis*_-,
                                               '.                 ,                 . ~o".JJ,rol~ing M~thQd ._-

11 Mechanic N/A N/A 12 Electncran N/A N/A 13 l&C Technician N/A N/A 14 Other N/A N/A 15 Other N/A N/A COOPER Page 33

COOPER ON-SHIFT STAFFING ANALYIS REPORT ~eBrigade 1 N/A N/A 2 NIA NIA 3 NIA N/A 4 N/A N/A 5 NIA NIA The LOCA does not include a fire. 0 COOPER Page 34

COOPER ON-SHIFT STA~tfNG ANALYIS REPORT J CNS.rAalE.4.:..'RADIATJON*PR0TECtlON AND CHEMIS,tRY*'*,:,'** , ,,' ,.J.:. -.:,*'-.<~.> < ;':}.?*'\::'; .. _* * *: Anal;ysi(ff3 Loss*OtC~lanfAcclden( *_ *:,. ,:. * ."_ ._*;*". .:_.*-*>-: --_ .. ~- ::*_. __ . L.. ,Position Performing

         !'. . i=unctioh* / Task-N ** .: .

E . -- - 1 In-Plant Survey* Set up portable sampler for CR X X X monitoring 2 On-site Survey: N/A 3 Personnel Monitoring NIA 4 !Job Coverage: Proc 5 2 Att 1 (Operator valve man1pulat1ons X X X X X X w/in 100 min of SD) 5 Offs1te Rad Assessment:

                'Included in Table 5) 6      Other site specific RP (describe). RP Briefing/                                         X     X      X     X     X Issue Kl 7      Chemistry Function task
                #1 (describe) - N/A 8      Chemistry Function task
                #2 (describe) - N/A
      *Performance Times are estimated Note Tasks are as directed by the SM. The only time critical task Is job coverage for operators to manipulate valves within 100 minutes of shutdown TSC and OSC are operational within 60 minutes to provide RP support.

COOPER Page 35

COOPER ON-SHIFT STAFFING ANALYIS REPORT

        '/f : -~- '!,,', _:,. ',>',~ :~ ,::' :e"1!$ TAE!l'.E'5/--:.'l=-10E~G,EN~.~~N '-~-i!LE~l:N!:AT;JON::--:* ~~                                      0:,-,   .. :- _'-:~}." '. *:;:::* -:,~ _,;

v, - -: .: * ,,*-: .- ., ' . c,* Anajysis#3LQS$.QfCoolant/\cbld'!Mlt *.. ~ ,, .*.:. * ~*'*_:-.:: * -*. '.':

          / ~ ).:_- /~~;;-~~a~*,::,,; ~ *~ *~~ ~ ,*_: c-, _*: ~ *
                                                     .r,-,.       h~- _, ' !~ * * ,~ *, * * : - ~ - ; ,; ::z 5i~ {~~~~ ~~~-~~,, *~ /-*_) *<:~; :~/;~:.~,i~ ~. * * ,*> ~ ~/p~* **F*,j~:; ~*:* :*
       '.l:.foii *.;,;_~-._-, . *";~*.,, 'Eunctloh"r/ Tnl&:.:: '. ':_ :--_r,:,** :: ,*;:!(Qi1¥Shift;jbs;tlon~ r;;:~t_asl{An~sls, #i~l I

Li( .: J>. ,-:.:* -~:*. :>~ '~: ?_: -~_:-' *- _~: ;~):, :,~- ,:_ ~~:~~.:*:~/:~:*.:~~~-:k/ :-~~~~~1~ii:~_;~ ~ Emergency Planning

                     !Declare the emergency classification level IS .

1 (ECL) h1ft Manager !Training Program / EP Drills

                     ~pprove Offsite Protectrve Action 2

Recommendations NIA NIA Emergency Planning 3 !Approve content of State/local notifications IShift Manager tTraining Program 4 Approve extension to allowable dose NIA NIA Licensed Operator Notif1cat1on and direction to on-shift staff Training Program / 5 e.g to assemble, evacuate, etc ) Shift Manager Emergency Planning trraining Program Emergency Planning 6 ERO notification Shift Manager trraining Program

                     !Abbreviated NRC notification for DBT 7

ievent NIA NIA

                     !Complete State/local notification form                                                                                 Emergency Planning 8                                                                                     Communicator Training Program 0           9         !Perform State/local notifications                                         Communicator Emergency Planning Training Program Non-Licensed Operator 10 !Complete NRC event notification form                                              STE Training Program Emergency Planning 11 !Activate EROS                                                                     Shift Manager trraining Program Emergency Planning 12 Offsite radiological assessment                                                    Dose Assessor h"raining Program Emergency Planning 13 Perform NRC notifications                                                          :Communicator trraining Program
                    !Perform other srte-spec1f1c event Licensed Operator 14 P1otificat1ons (e.g., Duty Plant Manager,                                          IShift Manager trraining Program INPO, ANI, etc.)
                                                                                                                                            !Security Training 15 Personnel Accountability                                                           ISecunty Program I EP Drills COOPER                                                                                                                                                                                             Page 36

COOPER ON-SHIFT STAFFING ANALYIS REPORT D. Design Basis Accident Analysis #5 - Fuel Handling Accident.

1. Accident Summary

The fuel handling accident is assumed to occur as a consequence of a failure of the fuel assembly hfting mechanism, resulting In the dropping of a raised fuel assembly onto the core. The drop causes mechanical damage to the bundle being dropped.
The drywell and reactor building are open with ventilation running. All releases to the environment are through the reactor building vent1lat1on.
The drop of a fuel bundle in containment over the reactor pressure vessel bounds a drop of a fuel bundle in the reactor bu1ld1ng due to the greater drop height.
Smee Secondary Containment Is not assumed to be functioning, the discharge is a ground level, unfiltered release from the ventilation exhaust plenum to the discharge point on the Reactor Building roof. This release point was determined to provide the most limiting dose consequences over other Reactor Building hatches, doors, and airlocks

2. Accident Specific Assumptions Made

Onsite personnel will be in accordance with the refuel outage staffing plan that includes additional SROs, ROs, NPOs, RP Techs, and Maintenance.

3. Procedures for Accident Response

EOP-5A, EOP Flowchart, Radioactive Release Control
5.1 RAD, Building Radiation Trouble
5. 7 .1, Emergency Classification

,')

5.7.2, Emergency Director EPIP
5. 7 .6, Notrf1cat1on
5 7 .16, Release Rate Determination
5.7.17, Dose Assessment
5.7.19, On-site Radiological Monitoring COOPER Page 37

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing l1:;

            *:~-..

r 0-n-s _-:-h.:ift;::::-'

   ~~~I~~:~-~;~~-~~:*:.~::,/                    - ~~~~r*-,r -~:~t~(: .-:;:::7:~r~~ *_:; t*L1~~,_,-,<,:
                                           *,.; . ,, . ,;_ ', ',,:, :*:.'*i/">'""l{gmeii~ion,*

0 R9IEi"inTable.#*,;u~.-:-:"-,j, ,:., .._<cTMS'_.-* T2/L 1

                                                                                                          ., ' I   nancJtyZ-.. ,        'I
                                                                                                                 *:~;~~ki ?_~: :k~u_l~ed:

T5/L 1 T5/L3 Emergency Plan 1 !Shift Manager 60 T5/L5 No Yes Figure 5 2-1 T5/L6 T5/L 11 T5/L 14 Control Room Emergency Plan 2 N/A T2/L2 No No Supervisor Figure 5.2-1 Shift Technical Emergency Plan 3 N/A T5/L 10 No No Engineer (STE) Figure 5 2-1 Reactor Operator Emergency Plan 4 N/A T2/L4 No No 1#1 Fiqure 5 2-1 Reactor Operator Emergency Plan 5 N/A T2/L5 No No 1#2 Fiqure 5.2-1 Reactor Operator Emergency Plan 6 N/A N/A No No 1#3 Fiqure 5 2-1

                   !station Operator                 Emergency         Plan 7                                                                             N/A          T2/L6              No                   No 1#1                            Fiqure 5.2-1
                   !station Operator                 Emergency         Plan           N/A           N/A 8                                                                                                            No                   No 1#2                            Fiqure 5 2-1 Station Operator                  Emergency Plan                   60 9                                                                                           N/A               No                  No
                   #3                             Figure 5 2-1 T5/L8 Emergency Plan 10          Communicator                                                       60           T5/L9              No                  No Figure 5 2-1 T5/L 13 RP/Chem Tech                     Emergency Plan                   60 11                                                                                          T4/L3               No                  No
                   #1                             Figure 5.2-1 Emergency Plan                   60          T5/L 12 12          Dose Assessor                                                                                      No                   No Fl!=)Ure 5.2-1 Utility Fire Brigade                          Emergency Plan                                 N/A                                    No 13                                                                            N/A                               No Figure 5.2-1 Member #1 Ut1l1ty Fire Emergency Plan                                 N/A 14           Brigade                                                          N/A                               No                  No Figure 5.2-1 Member#2 Security 15          Security                                                           60          T5/L 15              No                  No Continqency Plan COOPER                                                                                                                                         Page 38

COOPER ON-SHIFT STAFFING ANALYIS REPORT

             , .;* . ~ ... * :. ' . *CN&**,:-AB~E 2*_,_ 'PLANT. QPERAilONS ~ ~AF~ *SHUTD.O,W_N. . * .                                                            >,::. ~-* ::*. **:: ~ *
          ;. : *. '*_., >_:, ,*. *-,. **>: (. : : -~*: , *:.: .

dn'e Unit""" One Control Room .~- * .*.;,: * ; :' <' ./.:f, 1

                                                                                                                                                                               .;    1 :. ,-

.: ' '< -.. >* ' : :: ***::. _, _,: . . ANALYSIS*# s*FuefHandllriaA.ccldent **::/. > *: ,*_* .... ,**~'.'*, *.~ *X:>,

Mi~il)lum Operaffc;ins '!;:,few_ N~ssary1~)m13l~.meint A.OPs-~nd EOP_s, tjr*S~M~s }f Applicab,le *..:, ~ --- */ *'

          ~ ># *,~',   *~~*<'    ,*.,*I.--,,'_, ,-,.,**r,        ,::,Q,,      ,J"*f~:;.,, l;, *, *'     '*  *:Ct     '/:, *.   ~ ~*:   *'   if.*_,*,**. ,'~i ,'", ',:"', ,:.*--. *-~~~* .-:

Licensed Operator 1 !Shift Manager Shrft Manager rrraining Program r- Licensed Operator 2 Unit Supervisor '-'ontrol Room Supervisor rr . P raining rogram Shift Technical Engineer Licensed Operator 3 !Shift Technical Advisor (STE) rrra1ning Program Licensed Operator 4 Reactor Operator #1 Reactor Operator #1 rrra1rnng_ Program Licensed Operator 5 Reactor Operator #2 Reactor Operator #2 rrra1ning Program Non-Licensed Operator 6 Auxiliary Operator #1 !Station Operator #1 Training Program

,) 7 Auxiliary Operator #2 IN/A NIA 8 Other needed for Safe Shutdown IN/A NIA 9 Other needed for Safe Shutdown N/A NIA 10 Other needed for Safe Shutdown N/A NIA Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs if Applicable Line#~ !Generic Title/Role. * .' 'l On'-Shlft, Position

                                                                                                                                  ,  '      tfask Analysis.             : : *; *-.~-- -~
        ,  : -~ .:.\
                 .   -. ,f  * '*     :_:: -= ~ ~ -:.  * *,                                                     .,                                       1
                                                                                                                                          . K:ontrol 1i~g.M~thod_ *: : .
                                                           <,  *   : , 'T
                   -                ..                            .....  / ~-             ,

11 Mechanic NIA N/A 12 Electrician N/A N/A 13 l&C Technician N/A NIA 14 Other N/A N/A 15 Other N/A N/A COOPER Page 39

COOPER ON-SHIFT STAFFING ANALYIS REPORT 1 NIA NIA 2 N/A N/A 3 NIA NIA 4 N/A N/A 5 N/A N/A FHA does not include a fire. COOPER Page 40

                                                                           ,..,_ t...,

(_ COOPER ON-SHIFT STA~NG ANALYIS REPORT 2 On-site Survey: N/A 3 Personnel Monitoring: RP/CHEM#1 X X X X X 4 Job Coverage. N/A 5 Offsite Rad Assessment: (Table 5) 6 Other site specific RP describe): N/A 7 Chemistry Function task

             #1 (describe)

N/A 8 Chemistry Function task

             #2 (describe)

N/A

Times are estimated Note- RP/Chem#1 assists in monitoring and decon of building evacuees. Additional RP on shift for refueling/outage COOPER Page 41

COOPER ON-SHIFT STAFFING ANALYIS REPORT

       . "<;,        *. i,-,:.r,...
                            ';-'..' ,,,:_,J'(;=".~-                :**-*:cNs ~e::s:- e;,,E,JtG~9;t                                                  PUO.f ~BLEMEJtfATIPN                                 ';,:*.,: <:' *.,, *,;:,,:>; \

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                                                                                    !l,i"',.\~ ._.,,
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                                                                                                                      . ~.tfS!F 1
                                                                                                                           . - ue*1 ~,       A'an--"q-*1dll,1nra-.,.:,:~ccid&ir
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                                                                                                                                                                                  , . : *.:,~ *'!*.'~* '*,.~,.,

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                             ~, . --_, ...'*,.,***J~ .,. .. - F.,Ul)-Ct:     .*...... ~..,        1,:r-**-.k---.-..... *- .-r, ",;; ,:;:-:*:0n.:Slifftf?oslHon**.~"--

(i~J1# I,,,...,.,..,,._...,~: . .

       ?:: :._"": ~\ ~:.;~::~t[~:J(~. \.~-~/??i: ~~: ~~ ..:_;,~4'~:~~-~i::' ;: >

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                                                                                                                                ... ~~-~- *.,~.!. \~-;} :*~):/!}~-~ ~ ~:~ ::',~-: "-;.;\/: ~--;'JC6:ntto(llrig"Metlio
                                                                                                                                                                                                        -;;:; r~-~~ria)¥S1$*: *:-;*~

d*:':. Declare the emergency classification level Emergency Planning 1 Shift Manager rrram1ng Program / EP

                         ~ECL}

IDrills Approve Offsite Protective Action 2 N/A N/A Recommendations 3 !Approve content of State/local notifications !Shift Manager Emergency Planning rrraining Program 4 !Approve extension to allowable dose IN/A N/A Licensed Operator Notification and direction to on-shift staff rrra1ning Program / 5 !Shift Manager

e.g., to assemble, evacuate, etc.) Emergency Planning rrrainmg Program 6 ERO notif1cat1on Emergency Planning

                                                                                                                                         !Shift Manager rrrammg Program 7         ~bbreviated NRC notrfication for DBT event IN/A                                                                                                                        IN/A 8         Complete State/local not1f1cat1on form                                                                                                                                    Emergency Planning
                                                                                                                                         !Communicator rrraining Program 9        Perform State/local notifications                                                                                                                                          Emergency Planning Communicator rrra1ning Program 10        Complete NRC event notlf1cat1on form                                                                                                                                      Non-Licensed Operator ISTE
                                                                                                                                                                                               !framing Program 11        Activate EROS                                                                                                                                                             Emergency Planning
                                                                                                                                         !Shift Manager 1fra1ning Program 12         Offs1te radiological assessment                                                                                                                                          Emergency Planning IDose Assessor

[Trammq Program 13 Perform NRC notifications Emergency Planning

                                                                                                                                         !Communicator Training Program Perform other s1te-speclf1c event 14         not1f1cat1ons (e.g., Duty Plant Manager,                                                                                                                                  Licensed Operator Traininf
                                                                                                                                         !Shift Manager INPO, ANI, etc)                                                                                                                                                           Program 15         Personnel Accountability                                                                                          Security                                              Security Train mg Program COOPER                                                                                                                                                                                                                                                            Page 42

COOPER ON-SHIFT STAFFING ANALYIS REPORT E. Design Basis Accident Analysis #6 - Main Steam Line Break.

1. Accident Summary

This event involves the postulation of a large steam line pipe break outside containment downstream of the outermost isolation valve with a simultaneous loss of offs1te power. No fuel damage is calculated to result but the EAB dose assumes coolant 1-131 TS limit of

5 4 µCi/gm MS IVs isolate resulting In a steam cloud (puff) to environment within seconds.

2. Accident Specrfic Assumptions Made

Emergency Classrf1cat1on Is not required based on the break of the main steam line. This analysis is based on the FSAR MSLB accident dose. Assume the turbine building vent reaches the Alert level for~ 15 minutes (EAL AA 1 1)

3. Procedures for Accident Response

5.3BREAK, Pipe Break Outside Pnmary Containment
5 1 RAD, Building Rad1at1on Trouble
5. 7.1, Emergency Classification
5.7.2, Emergency Director EPIP
5 7 6, Notrf1cat1ons
5 7 16, Release Rate Determination
5.7.17, Dose Assessment
5.7.19, On-site Rad1olog1cal Monitoring

._____,,} COOPER Page 43

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing.

                                                      ..    ; ~-:Cl"$ TAaLE J-~ _(?N-SHl~*Posmo~:s -;                           ~     --         * ,l, :* <,      *
                                 .,    ', .         ;: ,*. ,f'n~ly~ls,tt §_ Main Sfeam Line Break {MSLB}                               0          t:J __ ,. -     ,
                     ..       ,,                                                                                                      J_   :;._ '_,i       - - "..., __ ..:   ....I-~ - -      - ......,._ - -
                                                                                  -~ --

Augmentation Role'.i~ Ta~I~ -- -* - ' -J--:-*~ - - -----:

    *une* On-shift                                                                                                                      Onan~lyzed                                  *TM,~ - *>
                                         ,   -    IE-Plan f{efei"ence *                 .,Elaps~ Time,. '>',#- /,Line      #  0
       # P0$1tloo                          
                                           .,   , --,<- ,    - ;r* ... , ~':.;.**-: , . :* -_ . '(inlnf -.* _        : ,,
                                                                                                                       -         1
                                                                                                                                   ,-i! _, -t~k?
                                                                                                                                                                             *.Required?
                               -  -                 :          -- -                     r          ** *  *   '.                    _.._ -   ,4'         ,     .,

T2/L1 T5/L 1 Emergency Plan T5/L3 1 !Shift Manager 60 T5/L5 No Yes Figure 5.2-1 T5/L6 T5/L 14 T5/L 11 Control Room Emergency Plan 2

                     !Supervisor                  Figure 5.2-1 N/A             T2/L2                            No                                    No
                     !Shift Technical             Emergency Plan                                                  T2/L3 3

Engineer (STE) Figure 5.2-1 60 No No T5/L 10 Reactor Operator Emergency Plan 4 Figure 5 2-1 N/A T2/L4 No No 1#1 Reactor Operator Emergency Plan 5 Figure 5.2-1 N/A T2/L5 No No r#2 Reactor Operator Emergency Plan 6 N/A N/A No No W-3 Figure 5.2-1 Station Operator Emergency Plan 7

                     #1                           Figure 5 2-1 N/A             T2/L6                            No                                   No Station Operator             Emergency Plan                                  N/A              N/A 8
                     #2                           Figure 5.2-1 No                                   No Station Operator             Emergency Plan                                   60              N/A 9
                     #3                           Figure 5 2-1 No                                   No T5/L8
                                                 !Emergency Plan 10              Communicator Figure 5 2-1 60             T5/L9                            No                                    No T5/L 13 RP/Chem Tech                 Emergency Plan                                   60             T4/L 1 11
                     #1                           Figure 5.2-1 No                                    No Emergency Plan                                   60            T5/L 12 12              Dose Assessor                                                                                                                 No                                   No Figure 5.2-1 Utility Fire Emergency Plan 13              Brigade Member Figure 5 2-1 N/A              N/A                             No                                   No
                     #1 Utility Fire Emergency Plan                                                   N/A 14              Bngade Member Figure 5.2-1 N/A                                              No                                   No
                     #2
                                                 !Security Contingency 15              Security                                                                     60             T5/L 15                           No                                   No Plan COOPER                                                                                                                                                                                                        Page 44

COOPER ON-SHIFT STAFFING ANALYIS REPORT tv,irifmuii"( ()perations* Grew,:N~ary'to *!mplerfl~iit' A~ps J~nd '~0Ps &- S:AMGs _lf1/2ppl~qle-**1 * ,., , * '

.,_:..... ' ' _ / :*.., **~- ,.**-' ,-*,***-: ~-~ ,*:,w,;, ,_**4- ___ ::.r.* :*,~.-: ,,_r," -~* ...._. * ~ . -***f* __ ; ' ~ , - - * * , , . ' *~ ,._

Licensed Operator 1 IShrft Manager Shift Manager Training Program Licensed Operator 2 Unit Supervisor Control Room Supervisor Training Program Shift Technical Engineer Licensed Operator 3 !Shift Technical Advisor (STE) Training Program Licensed Operator 4 Reactor Operator #1 Reactor Operator #1 Training Program Licensed Operator 5 Reactor Operator #2 Reactor Operator #2 Training Program Non-Licensed Operator 6 Auxiliary Operator #1 Station Operator #1 rTraining Program 7 Auxiliary Operator #2 N/A NIA N/A IN/A ~J 8 9 Other needed for Safe Shutdown Other needed for Safe Shutdown N/A N/A 10 Other needed for Safe Shutdown N/A N/A Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs if Applicable LlrnF# ~neric T~le!R?le - -. ' , o*n::Shift Position.*,

                                                                                     "                            ,,     , ,*   I T~k r An~ysis
                                                                                                                                                                    ~ , ,*
                                                                                                    ~
                    "       ..:. ,             -  '.                     -                   , . --- -      \                     C_Qi:,trollJ~g M~h~d* _

11 Mechanic N/A N/A 12 Electrician N/A N/A 13 l&C Technician N/A N/A 14 Other N/A N/A 15 Other N/A N/A COOPER Page 45

COOPER ON-SHIFT STAFFING ANALYIS REPORT

:* --, . *.,- *,.-:* .;* _-. *, . *-: *. "::**- :c.N~ TAeL~ 3 7 F:l~EFIGHTiNG '. **, .

                                                                                                                          , ' *r,,. ' -* ,
       .*:.,,,. :* -. *:; *_ ... ~ . . , -* -~ALY.SIS# 6Maln Steam Line Break (SLB\ *.: :.                                 ~*..,.: '
                                                                                                                                                - ,* ~.:::~ 'I, LIM .. *. , '*._ *, Pefforfned:by . _--. ._-*                              . :,_. . . 'Task/Analysis Cqfitrolllng Methoci                    . :"'* *.:
      *_/*:* ~ -~i ~:-.:~, ;._ *.>: _ ,~--~-- _* ~':.*_**~*;'_.'.-~**.:. ;:-_--:>-* ~' .. ,- -:_* __.-'.    . _.*:~-~<, . * .. :      }'? * ,:* :_'_;~*_:,:

1 N/A N/A 2 NIA N/A 3 N/A N/A 4 N/A N/A 5 N/A N/A Note- The MSLB analysis does not include a fire. COOPER Page 46

J {,,, COOPER ON-SHIFT STA~iING ANALYIS REPORT

                                                              . CNS TABLE 4 -:RADIATION:.PROTECTION AND CHEMIST>RY
                                                                      *, .' Analysis# 6 Main Steam Line Break '(MS*l:B\'." - * :*        .'
                                                                                                                                                     '.            '  .,.. I lin!3 # !Position Performing
                                                                                                                                                 *1.          .,

Perfo(mance Time Period After Emergency Declaration (minutes)* Funct,ion / Task ,, . ..... . ' ' . ' - ' . .. ': ~-' * -~A

~*

                    ~ ...._              't,,

6,5-

7b- 75-,; 80-

                            '        - *'                   0-5 5-10 1.0-15 15-20 20-2~ 25-30 130-35 135-40 140-45 145-50 50-55 55-60 60-65 7_5 - .80 , 85* * :85-90
                                 ,..,._._c **- *
                                                 ,_____,__                                                                                  - '         to 1   In-Plant Survey.

RP/Chem#1 Turb X X X X X X X X X Bldq 2 On-site Survey: N/A 3 Personnel Monitoring- N/A 4 Job Coverage. N/A 5 Offsite Rad Assessment. (Included in Table 5 6 Other site specific RP (describe) N/A 7 Chemistry Function task

                #1 (describe) N/A 8    Chemistry Function task
                #2 (describe) N/A
      *Performance times are estimated COOPER                                                                                                                               Page 47

COOPER ON-SHIFT STAFFING ANALYIS REPORT

     *'.* <~:!.\.:>':{ : .( .::f:" \).9N.s"!~BJ:;~_-!;;=~-~,,.E,~~1:NpJ1;f~!:(~*~M~~~TJQ~?'__ =-\*.'_ '.?/*/r:.:_,r.:::~ *,_::=.:.
       .~::,,.~.--~r;:~- _'. -*. * *,*:.::.(.~:,:*;Anafx~!s,*tl*~*Main-:?team:llne:Break)(MSLBt':
    '.:'~: :-:.-~:__.~:'~l:~--~-~ '/,\!:~ ::r;2:~~:f5~ ~~:': <c-:(r:;',{ ~. ~;,(~*~~~:y*,*~]~:~~; >'r~~- ~~ ~: -::~,::I!;~:_,~"~~*~~;~J. ---: ;*~~ ;~.i ---~---    **~=:~" , *. ',":'.**

_>_ .~ ,~r:;~~ .~, ~;--:~.: ,_(~~ /-': ~: ~~ ~;/;.~, ::~l

    ~ung,  ....

.:. :~r.~~~..-_;?,:Fun'ctk>n'ifTask,,.;-: _t;-~* ::-:*.r,¥.' ,*;*,r*on:ShtftRosttfon** ;* t :J:;ask~natysisfe . ' .... "--*~oltr***-

                                                                                                                                                                 ~-~--*Jig

_:, #_~; ?: -:~ ~/;~ *J '._ //* ~ -~/:<" <~~{ .; /; -=~\:*/ '_/',it~<-\ :: ;:~~i}< :)\:1 ~~ ~, ~ .~ -~J~;-::~~'1~ _.:.~ -:~_ -~ ~*:*

    ;,,~        ~
                                              ! //-/

Emergency Planning Declare the emergency classification level 1 Shrft Manager Training Program /EP ECL) Drills Approve Offsite Protective Action I 2 Recommendations N/A N/A 3 Emergency Planning Approve content of State/local notifications !Shift Manager Training Program 4 Approve extension to allowable dose N/A N/A

                                                                                                                                        ~1censed Operator Notification and direction to on-shift staff                                                                        !Training Program I 5                                                                            !Shift Manager

e g., to assemble, evacuate, etc.) Emergency Planning

                                                                                                                                       !Training Program 6        ERO notification                                                                                                      Emergency Planning
                                                                                       !Shift Manager
                                                                                                                                       !Training Program 7        lt\bbrev1ated NRC notification for DBT event                         NIA                                              N/A Emergency Planning 8        !Complete State/local notrf1cat1on form                               Communicator
                                                                                                                                       !Training Program Emergency Planning 9        !Perform State/local not1f1cat1ons                                   !Communicator
                                                                                                                                       !Training Program Licensed Operator Trainm~

10 !Complete NRC event notrf1cat1on form ISTE Program 11 Emergency Planning Activate ERDS Shift Manager ITram mg Program 12 Offs1te radiological assessment Emergency Planning Dose Assessor

                                                                                                                                       !Training Program 13 Perform NRC not1f1cations                                                                                                     Emergency Planning 1Commun1cator
                                                                                                                                       !Training Program Perform other site-specrf1c event not1f1cat1ons                                                                       Licensed Operator Tramm~

14 !Shift Manager e g., Duty Plant Manager, INPO, ANI, etc.) Program 15 Personnel Accountability N/A N/A COOPER Page 48

COOPER ON-SHIFT STAFFING ANALYIS REPORT F. Design Basis Accident Analysis #7 -Aircraft Probable Threat

1. Accident Summary

The analysis includes all emergency response actions taken prior to an aircraft impact in accordance with RG 1.214 for an aircraft threat that is greater than 5 minutes, but less than 30 minutes from the site, and considers the dispersal of the site fire brigade away from target areas for firefighting
The analysis does not include a scenano or response actions taken during or after a crash.

2. Accident Specific Assumptions Made

The Shift Manager receives the call from the NRC of probable aircraft threat.
All non-security on-shift personnel are inside the protected area fence at their normal workstation.

3. Procedures for Accident Response

5.5AIRCRAFT, Aircraft Threat COOPER Page 49

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing.

u-,. *; *.. i -* _.. *: -- . - -: ) : *, ';-. ':: r" '. *:-;i::- . *,

ne., ., * *.** . * * ~--*'"* .. *, Augnieritation. Role" lrfTable.*# u~ nana~""-'*

                                                                                                                  ~.,'1 * ~~- _- - TM 1 .. -s.-: *.- .',
    -_,~ -~~n~~~~-:~~-~~~-"-~;~~l~)~~~!l~ :: .~1~~:~~~e. _,'/Line.# ,:: -*~- -~~~? :*:* R~O~r~f T2/L 1 T5/L 1 Emergency Plan                                         T5/L3 1     Shift Manager                                                       60                               No                 Yes Figure 5 2-1                                            T5/L5 T5/L6 T5/L 14 Control Room                Emergency Plan                                          T2/L2 2                                                                        N/A                               No                  No Supervisor                  Figure 5 2-1                                           T5/L 11 Shift Technical             Emergency Plan                                          T2/L3 3                                                                         60                               No                  No Engineer (STE)              Figure 5.2-1                                           T5/L 10 Reactor Operator            Emergency Plan 4                                                                        N/A              T2/L4            No                  No 1#1                          Figure 5.2-1 Reactor Operator            Emergency Plan 5                                                                        N/A              T2/L5            No                 No 1#2                          Figure 5.2-1 Reactor Operator            Emergency Plan 6                                                                        NIA              T3/L1            No                 No 1#3                          Figure 5.2-1 7    !Station Operator            Emergency Plan Figure 5 2-1                           N/A              T2L6             No                 No 1#1
            !Station Operator            Emergency Plan                         N/A 8                                                                                         T3/L2            No                 No
            #2                           Figure 5.2-1
            !Station Operator            Emergency Plan                         60               T3/L3 9                                                                                                          No                 No l/t3                         Figure 5.2-1 Emergency Plan                                          T5/L8 10 Communicator                                                            60               T5/L9            No                 No Figure 5 2-1 T5/L 13 Emergency Plan                         60 11    RP/Chem Tech #1                                                                       N/A             No                 No Figure 5.2-1 Emergency Plan                         60 12     Dose Assessor                                                                         N/A             No                 No Figure 5 2-1 Utility Fire Brigade        Emergency Plan 13                                                                         NIA              T3/L4            No                 No Member #1                    Figure 5.2-1 Utility Fire Brigade         Emergency Plan 14                                                                         N/A              T3/L5            No                 No Member#2                    Figure 5.2-1 Security 15 Security                                                                60                N/A             No                 No
                                        !Contingency Plan COOPER                                                                                                                                            Page 50

COOPER ON-SHIFT STAFFING ANALYIS REPORT

                                                                                            -~       ..

CNS TABLE 2 - PLANT OPERATIONS & SAFE SHUTDOWN C i One Unit - One Control Room ANALYSIS # 7 - Aircraft Probable Threat Minimum Operations Crew Necessary to Implement AOPs and EOPs or SAMGs if Applicable Line# Generic Title/Role Pn-Shift Position Task Analysis Controlling Method Licensed Operator 1 Shift Manager Shift Manager rTraining Program Licensed Operator 2 Unit Supervisor Control Room Supervisor Ifraining Program Shift Technical Engineer Licensed Operator training 3 Shift Technical Advisor (STE) Program Licensed Operator

      ~          Reactor Operator #1                     Reactor Operator #1 Ifraining Program Licensed Operator 5          Reactor Operator #2                     Reactor Operator #2 rTraining Program Non-Licensed Operator 6         !Auxiliary Operator #1                   Station Operator #1 rTraining Program 7         !Auxi liary Operator #2                  N/A                      N/A 8         Other needed for Safe Shutdown           N/A                      N/A

) 9 Other needed for Safe Shutdown N/A N/A 10 Other needed for Safe Shutdown N/A N/A Other (n on-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs if Appl icable Line# Generic Title/Role On-Shift Position Task Analysis

                                              .,                                   Controlling Method 11      Mechanic                                            N/A                       N/A 12      Electrician                                         N/A                       N/A 13      l&C Techn ician                                     N/A                       N/A 14     Other                                                N/A                       N/A 15     Other                                                N/A                       N/A COOPER                                                                                                         Page 51

COOPER ON-SHIFT STAFFING ANALYIS REPORT ,,-- e Brigade

                   .   '                        CNS TABLE 3 - FIREFIGHTING ANALYSIS #7 Aircraft Probable Threat Line                  Performed by                        Task Analysis Controlling Method 1               Reactor Operator #3                            Fire Brigade Training 2                 Station Operator #2                           Fire Brigade Training 3                 Station Operator #3                           Fire Brigade Training 4                   Utility Worker #1                           Fire Brigade Training 5                   Utility Worker #2                           Fire Brigade Training FB relocates to outside the footprint.
  )

COOP ER Page 5 2

l COOPER ON-SHIFT sT~MNG ANALYrs REPORT I ' J L P.1 rt* "'P rf ~ *1 \) - ~ '*' ~ . ' ..........\,.' "~ *,, ~ ~.\-..;*.>>*:-:.,r-,-;;t,::. ~ ,-'v.r'"***,,\,. t ... -',.-~ ........ *'1...

          . :I - ,'..-~~ctl-'?(): - ;_ prkm rl$l:. * *., .' * .' -:. -.. '.'. -."*:..' ,
                  ~ r.Un       OA 1 *~S :*., . . . . ,.**_.,. ,. , ..*.-*.. -~<1_~i
                                                                                                         -Perfot.rn}m~.J.ime!Perf@d)i,f.ter.. Ei;rje}fuepcy,?Decl~~-on'-(rninutes)**::.:: ::*\,;.- ;,;\:~/1*:-~:,-**_ *_

_:

                                                                                                          *-::       :*.-,.,L..._,,: . ..t>*

1

                                                                                                                                                                                                                              ....... ,.,  -    .,_,r, ... _.., ... ,r"'" ---,_"**,,t * *
                                                                                                                                                          .~ .. ~*- :X::.'_; .,~., . . ~ ..... .;.;.~;*_,_._er.. ),:-- .-~.--*t**,*~"":}~**', r','"_'"- . . . ~-~... -*:.-. ...:i.r.,!'"'""~
           . NE..    ,  _'.,".?.: .. : :::-,<:~ ~:~'.* .. :i ;'_if           ~~1:0.fo.:~*51;20120~2E*l2s:'~~oi~ ~4'0 40;~{5 ~5!5~ so'i,~'";../ i~ ~~~:i }6,f~-~ iJP~:_; \*?fr,;/\~*<-~;~~

1

                             * *,

_,--=...i.:__ *.

In-Plant Survey.

                                                     ~        :* * ~'.~)__ ,   Jo'  **   ~*   *   ,. ***.,,/ ,. * ***...:, ~ - ("./-~          ' . , '_.*. t    ~~-*7 -~*' ~: ~ *~ * ,"_*/:'-~;

1,~$-l -~'fl'3,:, ,' :,-',~**<~) <~7 0~"~ ;,i_."19. : , :ruJ.,~,. ~~,~~\! *, ~1*, ".:

2 On-site Survey* N/A 3 Personnel Monitoring* N/A 4 IJob Coverage* N/A 5 i0ffs1te Rad Assessment

                     'Included in Table 5 6       Other site specific RP (describe) N/A 7      Chemistry Function task 1#1  (describe) N/A 8      Chemistry Function task 1#2 (describe) N/A RP/ Chemistry Tech goes with FB to location outside the footprint as directed by the SM.

COOPER Page 53

COOPER ON-SHIFT STAFFING ANALYIS REPORT

      ,. :-. -...". *.-.,_ ** '> *

CNS;TABLE:5*.:.. EMERGENCY PLAN IMPLEMENTATION. .c;. :--,, ~ *:; ..,, ', ,* -~ *

      . ?~ :* _/' .:},:*_-:".*,;:_*: . . '. -~:_' Anal~sis:#7 ~Aircraft Probable.Threat~~::-::-.~-*--:_. .-,: - ; /*. ~:~- ...- : *_-{::,..
      -u~ *----::;;--: - .. **\**:Function) Task .'.{;;.'      : ' : .~: ~ *. .' . ~On'-Stilft Posttltjn'.). : . Task-'AiJal~rs'.Cori~Jlipg

,--~~- -~- ?. -'-/, *. ?*~~:*: --~->~/. -~-, *, ___ .:i . _\*:-*_ .>: -:;_ :>~*_:/.?::* :,_ :.~.L~ -= -~-:~'-.:J~~~tfi9i*~~-.. ~_,.,: .*

1 Declare the emergency classif1cat1on level Shrft Manager Emergency Planning EGL) Training Program / EP Drills Approve Offsite Protective Action 2 Recommendations NIA N/A Emergency Planning 3 Approve content of State/local notrf1cat1ons Shift Manager Training Program 4 !Approve extension to allowable dose N/A NIA Licensed Operator Training INotrfication and direction to on-shift staff (e g., 5 Shift Manager Program / Emergency

               ~o assemble, evacuate, etc.)

Planning Training Program Emergency Planning 6 ERO notification iShrfl Manager Training Program 7 ~bbreviated NRG notification for DBT event IN/A N/A Emergency Planning 8 Complete State/local notrf1cat1on form !Communicator Training Program Perform State/local notifications Emergency Planning 9 !Communicator [Training Program 10 Complete NRG event notrf1cat1on form Non-Licensed Operator -:J 11 Activate EROS

                                                                              !STE K:RS Training Program Emergency Planning Training Program 12 Offsite rad1olog1cal assessment                                   NIA                               IN/A Emergency Planning 13 Perform NRG notrficat1ons                                         Communicator

[Training Program Perform other s1te-specrf1c event not1f1cat1ons Licensed Operator Training 14 Shrft Manager

               ~e g, Duty Plant Manager, INPO, ANI, etc)                                                         Program 15 Personnel Accountability                                          N/A                               N/A COOPER                                                                                                                                                  Page 54

COOPER ON-SHIFT STAFFING ANALYIS REPORT G. Design Basis Accident Analysis #8 - Control Room Evacuation and Shutdown

1. Accident Summary

Fire in the Main Control that requires control room evacuation.
If the nuclear system becomes Isolated from the main condenser, decay heat Is transferred from the reactor to the suppression pool via the relief valves. The incident detecbon circuitry initiates operation of the RCIC and HPCI systems on low water level which maintains reactor vessel water level, and the RHR suppression pool cooling mode Is used to remove the decay heat from the suppression pool if required. When reactor pressure fails below 100 pslg level, the RHR shutdown cooling mode is started.

2. Accident Specrf1c Assumptions Made

The ATC operator will perform the immediate actions of the procedure to initiate a manual scram and verify all rods in before evacuating the control room.
Subsequent actions are performed form outside the control room.

3. Procedures for Accident Response

5.1 INCIDENT, Srte Emergency Incident
5.4FIRE-S/D, Fire Induced Shutdown from Outside Control Room
5. 7.1, Emergency Classrfication
5.7.2, Emergency Director EPIP
5.7 6, Notification
5.4POST-FIRE-CONTROL COOPER Page 55

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables No te: T a bl e 1 contains the sta ffinq identi 1ed in s ection II.A, OS A Minimum s ta ffiinq CNS TABLE 1 - ON-SHIFT POSITIONS Analysis #8 - CR Evacuation & Alternate SD Line Augmentation Role in Table#

On-shift Position E-Plan Reference Unanalyzed TMS Elapsed Time I Line# (min) Task? Required? T2/L 1 T5/L 1 T5/L3 1 Shift Manager Emergency Plan 60 T5/L5 No Yes Figure 5.2-1 T5/L6 T5/L 11 T5/L 14 Control Room Emergency Plan 2 N/A T2/L2 No No Supervisor Figure 5.2-1 Shift Technical Emergency Plan T2/L3 3 60 No Yes Engineer (STE) Figure 5.2-1 T5/L 10 Reactor Operator Emergency Plan 4 N/A T2/L4 No No 1#1 Figure 5.2-1 Reactor Operator Emergency Plan 5 N/A T2/L5 No No 1#2 Figure 5.2-1 Reactor Operator Emergency Plan 6 N/A T3/L 1 No No 1#3 Figure 5.2-1

)       7 Station Operator         Emergency Plan N/A              T2/L6             No         No 1#1                      Figure 5.2-1 Station Operator         Emergency Plan                     N/A 8                                                                                 T3/L2             No         No 1#2                       Figure 5.2-1 Station Operator         Emergency Plan                      60              T3/L3 9                                                                                                   No         No 1#3                       Figure 5.2-1 Emergency Plan                                      T5/L8 10    Communicator                                                 60              T5/L9             No         No Figure 5.2-1 T5/L 13 Emergency Plan                                      T4/L4 11     RP/Chem Tech #1                                              60                                           No Figure 5.2-1                                                          No Emergency Plan                                       N/A 12     Dose Assessor                                                60                                No         No Figure 5.2-1 Utility Fire Brigade Emergency Plan 13                                                                                  T3/L4 Member #1               Figure 5.2-1                        N/A                                No         No Utility Fire Brigade Emergency Plan 14                                                                 N/A              T3/L5             No         No Member#2                Figure 5.2-1 Security 15     Security                                                     60             T5/L 15            No         No Contingency Plan COO PER                                                                                                                      Page 56

COOPER ON-SHIFT STAFFING ANALYIS REPORT CNS TABLE 2- PLANT OPERATIONS & SAFE SHUTDOWN One Unit - One Control Room ANALYSIS # 8 - CR Evacuation & Alternate SD Minimum Operations Crew Necessary to Implement AOPs and EOPs or SAMGs if Applicable Line# Generic Title/Role On-Shift Position Task Analysis

                                    - - ..                                         Controlling Method 1                                                                        Licensed Operator Shift Manager                                   Shift Manager      !Training Program 2                                                                        Licensed Operator Unit Supervisor                          Control Room Supervisor ITraining Program 3                                              Shift Technical Engineer Licensed Operator Shift Technical Advisor                             (STE)          !Training Program 4                                                                        Licensed Operator Reactor Operator #1                         Reactor Operator #1    !Training Program 5                                                                        Licensed Operator Reactor Operator #2                        Reactor Operator #2     !Training Program 6                                                                        Non-Licensed Operator Auxili ary Operator #1                      Station Ope ra tor # I ITraining Program 7    Au xiliary Operator #2                                N/A          N/A 8    Other needed for Safe Shutdown                        N/A          N/A
)         9    Other needed for Safe Shutdown                        N/A          N/A 10    Other needed for Safe Shutdown                        N/A          N/A Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs if Applicable Line# K;eneric Title/Role                        On-Shift Position          !Task Analysis Controlling Method 11    Mechanic                                              N/A                       N/A 12    Electrician                                           N/A                       N/A 13    l&C Technician                                        N/A                       N/A 14    Other                                                 N/A                       N/A 15    Other                                                 N/A                       N/A COOPER                                                                                                    Page 57

COOPER ON-SHIFT STAFFING ANALYIS REPORT 1 Reactor Operator #3 Fire Brigade Training 2 Station Operator #2 Fire Brigade Training 3 Station Operator #3 Fire Brigade Training 4 Utility Worker #1 Fire Brigade Training 5 Utility Worker #2 Fire Brigade Training COOPER Page 58

COOPER ON-SHIFT STA"f-ilrNG ANALYIS REPORT ( J

                       . ::* , _,-.>, ,. '*   'h -~ .. ~~¢~.l~, TABLE,,4-.,~!J~TI,9tfP.R©JE.CTl0t-J1'~p.~~lil;,~1$tRY~}/.:* ...-~:~~ *::.,:-\~;;:*, -. /**:::t;":'*:~'?3:' ,;',_,:'<

_ , ,;-:,. * ', * '* ~ ... -:,_:,, *,_, ~- ,*-::.-~:Analysls"~.; "GRfE:y._ir.m,tin"ri 8{AltEirri~fAiSO:-*,,~.,:,/* ':"-,-*i~;,:: *,?,<'-.**~ \:.~>*.'/ , >~ ~.,_ *, :,.; t ,;,:_, '. /.;~; 0 r, 2 On-site Survey: N/A 3 Personnel Monitoring: N/A 4 !Job Coverage: RP/Chem#1 Support FB X X X X X X X X X X X X X X X X X 5 Offsite Rad Assessment: 1(/ncluded in Table 5 6 Other site specific RP (describe): N/A 7 Chemistry Function task 1#1 (describe) N/A 8 Chemistry Function task 1#2 (describe) N/A COOPER Page 59

COOPER ON-SHIFT STAFFING ANALYIS REPORT _, _-, -~N-~: Tj\ij_l;E ~ :--:EN!ERGENf'f: ~1/2NJMP!,.~~~~Tf'. TION _. *_,

       .                                           ,. _, _*

Analysis #8-'7 CR Evacuation* & Alternate SD * -:

     'L(rie                         ,

F~~~ti~;m'/?~k

                                                                             , .. - *. ~ -_-- -Oh.:Shlft Posi~lon: PJ:ask Ana_lysls Contfollir:rg
                                             -- - .         - ---  ,; .r
                                                                               ...r - - - --~ ..... 4-r J'.L  ~--
                                                                                                                   -*---~_.,_  ,

Method

                                                                                                                                                            * -- ~ -

Declare the emergency classification level Emergency Planning 1 Shift Manager rrraining Program / EP ECL) Drills Approve Offsite Protective Action 2 Recommendations N/A N/A Approve content of State/local notifications Emergency Planning 3 Shift Manager

fra1ning Program 4 Approve extension to allowable dose N/A N/A Licensed Operator Training Notification and direction to on-shift staff (e.g.,

5 IShrft Manager Program / Emergency

              ~o assemble, evacuate, etc.)

Planning Training Program Emergency Planning 6 ERO notification !Shift Manager Training Program 7 Abbreviated NRC notif1cat1on for DBT event N/A N/A Complete State/local nottficat1on form Emergency Planning 8 !Communicator Training Program 9 Perform State/local notifications Emergency Planning

                                                                                                    !Communicator Training Program Licensed Operator Training 10      Complete                 NRC event notification form                                  !Communicator Program 11      Activate EROS                                                                                                        Emergency Planning
                                                                                                   !Shift Manager Training Program 12      Offs1te rad1olog1cal assessment                                                      NIA                             N/A Emergency Planning 13     Perform NRC notif1cat1ons                                                             !Communicator Training Program Perform other site-specific event notifications                                                                      licensed Operator Training 14                                                                                          IShrft Manager (e g., Duty Plant Manager, INPO, ANI, etc.)                                                                          Program 15     Personnel Accountability                                                               N/A                            N/A COOPER                                                                                                                                                            Page 60

COOPER ON-SHIFT STAFFING ANALYIS REPORT H. Accident Analysis #9 - Station Blackout

1. Accident Summary

At power, normal lineups and no add1t1onal events
HPCI and RCIC start to restore level 2 Accident Specific Assumptions Made
Assume the Shrft Manager recognizes power cannot be restored within 15 minutes after the loss before the 15 minute SBO EAL period has expired.

3. Procedures for Accident Response

5.3SBO, Station Blackout
2.4FPC, Fuel Pool Cooling Trouble
5.3ALT-STRATEGY, Alternate Core Cooling M1tigatmg Strategies
2.2 99, Supplemental Diesel Generator System
2.2.100, SAMG Diesel Generator System
0 39.1, Fire Watches and Fire Impairments
5. 7 1, Emergency Classlf1cat1on
5.7.2, Emergency Director EPIP
5 7 6, Notification
5.7.10, Personnel Assembly and Accountability

':) COOPER Page 61

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing.

               *       "    ' >o***:.     ,, ,*, ,,   ;*;'CN°"s'TABtEf:..:0N:.SHIFl""POSITIONS"--:, "":**:\"'* ',';"'-; *;',                 n*   ,, <,*,
      . ;_       , ** ,. ... ,. __ -*      ---~- *-*:*. __ :),ri_ah,rsi~_.#tf~~Station*s1ackout_>.: ___ -~ _.: *:: .. :~-/~ ::, **      _~-*-    :!
    *L: - . . - ., * : - . *: ,., - ** -, . ~ ~; : At.fgrnentati_on,., *Role:ln Tijble _;*Unan}ilyz~tf-- i ._

Tt-,s.-; --*:--

       '.1#8, o'n.:shfft Posl~on.,. 'aaJ;i~ Df?AUm~n\ - ~!~~-~ -tfTe, *,,:.sf ~lnE!'~--- \ . ~---: *l.;i_Sk?.- *:-: . . ':ReC:f~l~eil{ ,*..
                   ,           __             .. _*_:    ..        :., __ **_(m10):;: *_; *

_ - ,* . _____ : -~,, _,_::. "-* ." .~. -->_.--_-;_-~

T2/L 1 T5/L 1 Emergency Plan T5/L3 1 !Shift Manager 60 T5/L5 No Yes Figure 5.2-1 T5/L6 T5/L 11 T5/L 14 Control Room Emergency Plan 2 N/A T2/L2 No No

              $upervisor                     Figure 5.2-1 Shift Technical                Emergency Plan                                        T2/L3 3                                                                         60                                   No                  No Engineer (STE)                 Figure 5.2-1                                         T5/L 10 Reactor Operator               Emergency Plan 4                                                                        N/A               T2/L4               No                  No
              #1                             F1qure 5 2-1 Reactor Operator               Emergency Plan 5
              #2                             Figure 5 2-1 N/A               T2/L5               No                  No Reactor Operator               Emergency Plan 6
              #3                             Fiqure 5.2-1 N/A                 N/A               No                  No Emergency Plan 7     !Station Operator #1 Firi                                           N/A               T2/L6 I:,iure 5 .2 - 1                                                         No                 No 1c      .                       Emergency Plan 8     1vtatIon Operator #2 F" Igure 5 .2 - 1 N/A                 N/A               No                 No 9      Station Operator #3              Emergency Plan                                         N/A 60                                   No Figure 5.2-1                                                                                  No T5/L8 Emergency Plan 10      Communicator                                                         60               T5/L9               No                 No Figure 5.2-1 T5/L 13 11      RP/Chem Tech #1                 Emergency Plan                       60                N/A Figure 5 2-1                                                               No                 No Emergency Plan 12      Dose Assessor                                                        60                N/A                No                 No Figure 5 2-1 Ut1l1ty Fire Brigade            Emergency Plan 13 Member #1                      Figure 5.2-1 N/A                NIA               No                  No Ut1l1ty Fire Brigade Emergency Plan 14                                                                          N/A                N/A               No                  No Member#2                       Figure 5 2-1 Security 15      Securrty                       Contingency                          60               T5/L 15             No                  No Plan COOPER                                                                                                                                              Page 62

COOPER ON-SHIFT STAFFING ANALYIS REPORT 1 Shift Manager Licensed Operator Shift Manager Training Program 2 Unit Supervisor Licensed Operator Control Room Supervisor Training Program Shrft Technical Licensed Operator 3 !Shift Technical Advisor Engineer (STE) Training Program 4 Licensed Operator Reactor Operator #1 Reactor Operator #1 Training Program Licensed Operator 5 Reactor Operator #2 Reactor Operator #2 Training Program 6 Auxiliary Operator #1 Non-Licensed Operator Station Operator #1 Training Program 7 Auxiliary Operator #2 NIA NIA 8 Other needed for Safe Shutdown NIA NIA

--)

-~

9 Other needed for Safe Shutdown NIA NIA 10 Other needed for Safe Shutdown NIA NIA Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs if Applicable Llntil#:-'.. Generic Title/Role' , " K:>n-Shiff Pd~~on , .

                                                                                        ....  ,r*
                                                                                                 . Tas!f-A(lal~is* , - '- -
                                                                        '           ,_    ,,, : ~11_!r6l~~g Metl_lod :

11 Mechanic NIA NIA 12 Electrician NIA NIA 13 l&C Technician NIA NIA 14 Other NIA NIA 15 !Other NIA NIA COOPER Page 63

COOPER ON-SHIFT STAFFING ANALYIS REPORT Fire Brigade 1 NIA NIA 2 NIA NIA 3 NIA NIA 4 NIA NIA .::) 5 NIA NIA The Station Blackout Analysis does not include the need for firefighting COOPER Page 64

COOPER ON-SHIFT STA~JNG ANALYIS REPORT CNS* TABL.:E 4 :- RADIATIP,N*:eROTECT.IONi AND~CHEM!STRY

           ',             >        ..           .       '      ;:., ,    ..,. *1     *,     r      '-  , ,.-. *     ).   ,* *  ~   'i-.-.    --;          <'      ,   -      ,       -              -  *           ..                          *         - --

__ . _ '* ,. .* - .~ "' ... , **;

                                                                   , . _ * -- c. *.. ~Analysis' #9 - Station *Blackout
                                                                              \'      ... ,     ._<..:*r*'       !-:,, \, '~**
                                                                                                                                                                                      . . --~- - *, *                       . ,. - .       '*                                 e
                                                                                                                                                                                                                                                                                                      - - -~          :

L IPosltion Performing** ' C

, ,,"!-,., ,', .~...'":,, ~.fU *::,.,,.. > :** ~ .. i , * *"" \',. J_)* L'\J"'* .... -I :, * ,. *"

                                                                              -      *: Perfomi'ance Time Period Afte}:6:merg~rcy*Declaration (i:ninutes')*.                                                                                                                    -.*                                            ,I I      Funct,ion / )"ask"                                                                                                            'l,,     *:,          ' .,....                  * *    . '*                                              ~                                                      .. ,..
                                                                                                                                                                                                                                                                                                               ,*r;,
                                                                                 -.,    *     -' -                j                   <  ','                                                                  ,, .;                                                       ~*I'         * *
       ,N

_5-:10, 10-15 '65; ' , 10'-' 75c: ,

E- ,.

I l Q;-5 .. ' 15-20 120.,,2~ l25-30 130-35 l35-4Q :40-4_5 45-50 50-'55 ~.5~.o 60.:S5 '90,

                                                                                                                                                                                                                           -'  ..,          *:.        :70~               75.-, '80'           '... *~--~--
                                                                                                                                                                                                                                                                                                    ~                  . ,,'

1 In-Plant Survey: N/A 2 On-site Survey. N/A 3 Personnel Monitoring. N/A 4 IJob Coverage. N/A 5 Offs1te Rad Assessment. 1

                  /nc/uded in Table 5-N/A 6      Other site specific RP N/A 7     lchemistry Function task 1#1: N/A 8      Chemistry Function task 1#2 (describe) - N/A RP/Chemistry Tech does not have an assigned task for SBO.

COOPER Page 65

COOPER ON-SHIFT STAFFING ANALYIS REPORT

    . ,.: / .., : _- ;.:.. _. : *:* - :.~, ::*~N?*;r,~ai:EJ,;..:E~J~R~eN*c;;v. *~.L-A!f rtv1PL~~'=NTAi:10N_                                                                             -
                                                                                                                                                               '.:*-* .:.~*- :,: ~;*,_:_,'.,:/:.*
     . "_.... *.* . */** ~* .. ~,--" -*. ',,*, . * *: * : : ~Analysis* #9 ...:station Blackout* - *_ , -' *                                                                                        ..
       -   ,, -,_,;__ -   rn; > -  , - - ....l,.:, ~ -* ~ -- ~ ' -     -  .-_ * ..:..* - _.,.. 'il."      -                                    -         * --  .. - ,.                  -*--" -
     -_una_ ,-. ** *.~ * - ..- : *,F'l.,nct1o*n.l'Task~<- - __ * *, :. ., .:

i, ". -. , -'O~hift Pqsi!ion; .. Task AnalY,S!_~ Con'tr911ing

    ~ # :,
     -**-. '            ::,., ~-: -;__-.~* :_\:_l               --.-:(<"*~-;J_*:.-:: .-:.- *:.:       .*,   ~ : .. :~,.- *.~,~---' ~- * ,"' :~   :* f   ..... *_., Methbt(*:_*-:~.:_ :*_ ':

beclare the emergency classif1cat1on level Emergency Planning 1 Shift Manager rrraining Program / EP ECL) Drills

                        ~pprove Offs1te Protective Action 2

Recommendations N/A N/A 3 Approve content of State/local notifications Emergency Planning Shift Manager Ifraining Program 4 Approve extension to allowable dose N/A N/A licensed Operator Notification and direction to on-shrft staff rTrammg Program I 5 Shrft Manager (e.g., to assemble, evacuate, etc ) Emergency Planning ITraming Program 6 ERO notification Emergency Planning Shift Manager rTraining Program 7 !Abbreviated NRC notification for DBT event N/A N/A 8 !Complete State/local nobfication form Emergency Planning Communicator Training Program 9 Perform State/local notifications Communicator Emergency Planning rTra1ning Program Licensed Operator 10 !Complete NRC event notification form STE

                                                                                                                                                      !Training Program
                      ~cbvate EROS                                                                                                                            Emergency Planning 11                                                                                            Shift Manager                                    Training Program 12             Offs1te radiological assessment                                                N/A                                               N/A Emergency Planning 13 Perform NRC notifications                                                                   Communicator                                     rTra1ning Program Perform other site-specrf1c event not1f1cat1ons                                                                                 Licensed Operator 14                                                                                            Shift Manager e g , Duty Plant Manager, INPO, ANI, etc.)                                                                                    rTraining Program 15              Personnel Accountability                                                       Securrty                                         Security Training Program COOPER                                                                                                                                                                                             Page 66

COOPER ON-SHIFT STAFFING ANALYIS REPORT I. Accident Analysis #10 - LOCA/General Emergency with Release and PAR

1. Accident Summary (Assumed for Staffing Analysis Purpose)

The unit is ,n a Site Area Emergency AS1 when the Shift Manager is given a dose assessment update that projects >1 Rem TEDE dose at the site boundary.

2. Accident Specrf1c Assumptions Made

All actions for SAE are complete.
No transients other than LOCA are considered.
The ERO would be activated at an Alert or SAE For Staffing Analysis purpose, the T=O clock Is used for the emergency plan actions to evaluate the capability to implement the GE class1flcatJon, PAR and notification functions before the ERO arrives.

3. Procedures for Accident Response

5.7 1, Emergency Classification
5.7.2, Emergency Director EPIP
5 7 6, Notification
5. 7 16, Release Rate Determination
5 7.17, Dose Assessment
5 7 20, Protective Action Recommendation COOPER Page 67

COOPER ON-SHIFT STAFFING ANALYIS REPORT

4. Tables Note: Table 1 contains the staffing identified in Section II.A, OSA Minimum Staffing.

                                                    *,: *":****:-~cNs tABUH'.:: bN-SHIFTP'osfTIONS*. '.,: *..-;*'.. :*.,              , .-- , -. -~ -~ ....
                         - ---- _~,.  ~*,:..~     ::.~-,::_: <:'._--A.n.alysjs:#10*:-LQCA/GE.with:PAR:_:~':.~ \_..-_:, . .:,:-;.* -- -* ,. ---- ,, ,

T2/L 1 T5/L 1 Emergency Plan T5/l2 1 !Shift Manager 60 T5/L3 No Yes Figure 5 2-1 T5/L4 T5/L5 T5/L 14 Control Room Emergency Plan 2 N/A T2/L2 No No Supervisor Figure 5 2-1 Shift Technical Emergency Plan T2/L3 3 60 No No Engineer (STE) figure 5.2-1 T5/L 10 Emergency Plan 4 R~ctor Operator #1 N/A T2/L4 No No figure 5 2-1 Emergency Plan 5 Reactor Operator #2

_'~) !Figure 5.2-1 Emergency Plan N/A T2/L5 No No 6 Reactor Operator #3 N/A N/A No No Figure 5.2-1 Emergency Plan 7 Stab on Operator #1 NIA T2/L6 No No FiQure 5.2-1 Emergency Plan N/A N/A 8 !Station Operator #2 No No figure 5.2-1 Emergency Plan 60 9 Station Operator #3 N/A No No Figure 5.2-1 Emergency Plan T5/L8 10 Communicator 60 T5/L9 No No Figure 5.2-1 T5/L 13 Emergency Plan 11 !RP/Chem Tech #1 60 N/A No No Figure 5.2-1 Emergency Plan 60 T5/L 12 12 Dose Assessor No No Figure 5.2-1 Utility Fire Bngade Emergency Plan 13 NIA N/A No No Member #1 Figure 5.2-1 Utility Fire Brigade Emergency Plan 14 NIA N/A No No Member#2 Figure 5.2-1 Secunty 15 ISecunty 60 T5/L 15 No No IContinQency Plan

,J COOPER Page 68

COOPER ON-SHIFT STAFFING ANALYIS REPORT

            ,-,.-_:".-  '.' *: '~:- *; __ :\~N~:.TA~fE?.:.,,Pl.fNT*Of:>>,l;~i;JP~S:*~.~"'F~-~f.lUTDPW~:, .,_,. '~                                                                        -.
          * ,, *           *          * * :;* ,*.. _           -*- ,_ . - .0ne'Untt .a.,:One Control Room - - *., *                                                                            ,.
                 .,:, :- *->.'*:,,.:_            .>:, * ,:~*: :,_-.;-          AAalysjs#1Q-~*LQCNGEwith~P.AA ** *1 .:*                                                                 _ --* ,.

rytiriim~m :Opei:af~ris*.GrEr,N *~~ to-1,rn~rneiif AQP~ ~Fld. EOPs'*or..?~M~s-if Applicable'. * * ,* =.,

         ;',~ ,          -., -".:~ ,,/:* *, ;- - -.*, . : ;=..' :.,,    r - ~  -  ,. -~*;_,  *,  __ , __ .t*,     .:  * ,;._ *:.-, :.,_,,_*',. __ - __ *.__:   ,- , ,,.    ,     , , j~~     ,- .,

1 Shift Manager Licensed Operator Shift Manager rrraming Program 2 Unit Supervisor Control Room Supervisor Licensed Operator fTraining Program Shift Technical Engineer Licensed Operator 3 Shift Technical Advisor (STE) fTrairnng Program 4 Reactor Operator #1 Reactor Operator #1 Licensed Operator fTraming Program 5 !Reactor Operator #2 Licensed Operator Reactor Operator #2 Jrainmg Program 6 !Auxiliary Operator #1 Non-Licensed Operator Station Operator # 1 Training Program 7 !Auxiliary Operator #2 NIA NIA 8 !Other needed for Safe Shutdown

) 9 Other needed for Safe Shutdown NIA NIA NIA NIA 10 bther needed for Safe Shutdown NIA NIA Other (non-Operations) Personnel Necessary to Implement AOPs and EOPs or SAMGs rf Applicable Line# * ~neric T~!e/Role_ - -- **'

                                                                                            - '  On-Shift Pc:>~itio_n :                                   . Ta;sk A~alysls:
                                                      '~*- ~ ,      , ,
                         --         )               ,
                                                                            ,, '                        ' ,.  *~ ~,
                                                                                                                                            -            : Controlling. MEith<>d*            ,

11 Mechanic NIA NIA 12 Electrician NIA NIA 13 l&C Technician NIA NIA 14 Other NIA NIA 15 Other NIA NIA l J COOPER Page 69

COOPER ON-SHIFT STAFFING ANALYIS REPORT 1 NIA NIA 2 NIA NIA 3 NIA NIA 4 NIA NIA 5 NIA NIA Note: The LOCA/GE with PAR Analysis does not include the need for firefighting. 0 COOPER Page 70

L COOPER ON-SHIFT STA't--1~NG ANALYIS REPORT

CNS TABLE 4 - RA:DIATION.,PROTECTION AND CHEMISTR Y : * "' . -"

       . ""                                              .. -' _

Analysis #10 -*LOCA.lGE wftfr.PAR: * -:. *- '-* - '.:* ** 0 l.'.! *- ' -

                                                                                                                                                                                                                                  "     '    r,:            *,

L Position Performing,_ : .;, :. .. ' . ;,;t','" .. ,., ... , . . . ., . ; *,

                                                                                                                                                                                                                                       \  ,,,,.,:::
      '     i *function/ task * *                                   Performanc e Jinie Period'After l;m~rgency.* Deelarat1on (minutes)**                                                                      ',          "'   .

N

                  . .                  I
                                                                        < '-     ~ '-     r

I , * * * '

l ._,

                                      -.                                                                                                                                                              70- t~ : 75_* : ab~- ,.,_._* .

E'* .. '0-s5 ~19* 10-15 15"-_~Q 120;-25 l25-30 39::~~ ~5~ 40~5 .- 14&-50 5()::55 ~5-60 ,66--_

                                                                                                                                                                               ~-~ ': ;70*.:                                                :as-~cf 7fJ' . .~'ao_, ~'as~;
                                                                                                                                                                                                         ' ** t

_ __ - ( 1 In-Plant Survey N/A 2 Ion-site Survey: N/A 3 Personnel Monitoring: N/A 4 Job Coverage. N/A 5 Offsite Rad Assessment :

               'Included in Table 5 N/A 6    Other site spectfic RP (describe) N/A 7   !Chemistry Function task 1#1  (describe) - N/A 8    Chemistry Function task 1#2 (describe) - N/A
      *Times are estimated. Note. The Dose Assessor is available for dose assessment COOPER                                                                                                                                                                    Page 71

COOPER ON-SHIFT STAFFING ANALYIS REPORT

          < *,~-: -~, *-,:.*. *,-; *.' *;,~- . ,CNS TABLE 5 - EMERG!=.NCY PLAN.l~Pl,.EM~NTATJON--..                         -      < -- -~, _
          -'_-::,,:*:   _,,      :,,:*~_.,.---:"     < Analysis#10-_LO_GA/(;Ewit_hRAR*;_*_,,>.:'_-_,*,*,l-                       0 LJn~ .~ : .;, * ~ *:-~, F.unctlon l Task~. ,:. - , ... _*

On~~lilft*P~ltion"" -.. -_; -'._* T~k ~rialY,s~* .: _, <

          . l . ~ - ::- ~----:~::*\ _;~~- :'*~- _.._--, ,,_*   : ~-. *.... -*-   :_ ,_-,    r,,         -~ * - CbrjtrollirigMettto~

Declare the emergency classification level Emergency Planning 1 !Shift Manager rTraining Program / EP ECL) Drills Approve Offsite Protective Action Emergency Planning 2 !Shift Manager Recommendations rTraining Proqram 3 Approve content of State/local notrf1cat1ons Emergency Planning

                                                                               !Shift Manager rTraining Program 4       Approve extension to allowable dose                                                     Emergency Planning
                                                                               !Shift Manager ifraining Program Licensed Operator Notrf1cat1on and direction to on-shift staff                                           rTraining Program /

5 !Shift Manager

                    ,e g, to assemble, evacuate, etc.)                                                     Emergency Planning Training Program 6      IERO notification                                           NIA                         N/A 7      ~bbrev1ated NRC notif1cat1on for DBT event IN/A                                         N/A 8      !Complete State/local notrf1cation form                                                 Emergency Planning
                                                                               !Communicator Training Program Perform State/local notrf1cat1ons                                                       Emergency Planning 9                                                                  Communicator if raining Pro!=lram

,-----._ Licensed Operator 10 Complete NRC event notif1cat1on form ISTE _) 11 Activate EROS NIA Training Program NIA Emergency Planning 12 Offs1te radiological assessment Dose Assessor Training Program Perform NRC notrf1cat1ons Emergency Planning 13 !Communicator Training Program Perform other s1te-specrf1c event notif1cat1ons iShrft M Licensed Operator 14 e.g, Duty Plant Manager, INPO, ANI, etc.) anager Training Program 15 Personnel Accountability N/A N/A COOPER Page 72

COOPER ON-SHIFT STAFFING ANALYIS REPORT VIII. APPENDIX C - TIME MOTION STUDIES SUPPORTING THE STAFFING ANALYSIS A. TMS Shift Manager Activate D1alogics COOPER TIME MOTION STUDY OF OVERLAPPING TASKS TASK 1: ACTIVATE THE ERO USING DIALOGICS JOB: SHIFT MANAGE R TASK 2: PERFORM EMERGENCY DIRECTION AND CONTRO L JOB: SHIFT MANAGE R COOPER Page 73

COOPER ON-SHIFT STAFFING ANALYIS REPORT PURPOSE: Perform a Time Motion Study to evaluate whether assigning the performance of ERO notification using DIALOGICS to the Shift Manager is an acceptable task overlap to the Shift Manager's primary emergency plan function of emergency direction and control. NOTE The Time Motion Study may be completed during simulator training/evaluation or during an EP drill while the Shift Manager is performing the Emergency Director function. LOCATION: Simulator REQUIRED TOOLS/EQUIPMEN T: A. Telephone with ANS Activation Module B. Procedure 5.7 2 Attachment 1 Activation of ANS-Hardcard C. Simulator Dialogics Communication Cross-Tie is in OFF or not cross-tied C) COOPER Page 74

COOPER ON-SHIFT STAFFING ANALYIS REPORT

 )PEND IX D (NEI 10-05)

Function/ Responsibility (Task) Analysis Template Even. Site: CNS Position: Shift Manaoer Line# 1 Function Responsibility (Task) Action Step Duration Clock/Step Time In 1.NotificatJon 1.1 Initiate notification to the CAUTION: ERO via the Dialogics The Simulator Communication Cross- tie for Program Dialogics must be OFF to prevent inadvertent activation of the ERO. +5/5 sec 11.1 Retrieve the Dialogics Hardcard InstructIon 1.1 2 At the ANS ACTIVATION MODULE, pick up handset +10/5 sec and verrfy dial tone. 11.3 +20 /10 sec Depress "ACTIVATE DIALOGICS" pushbutton. 1.1.4 When asked for Scenario, depress desired Yellow" +36/16 sec scenario pushbutton. 11.5

                                                                                                                             +48/12 sec Wait until system says "Goodbye" then hang up.

1.1 6 When pagers activate within 3 minutes, then STOP 1 :29 / 41 sec

                                                    ,END Of ERO ~OTIFICA~ TASK                                                               -   ~-*-c..~ "     .  ~ '
                                                                                                                   -~ '1 ,'*m1n, 29'sec>  .,, .,.,.--,- - - .:
                                                                                                                  "* r     -      ,>         *

f

                                                   -TOTAL; T ~ DURATIQN:                                                 ;
                                                                                                                                , , -'-,r~ "         *
                                                                                                                                                           ) .,
                                                                                                                                                                     ~
                                                                                                                                                                       ~

Function Responsibility (Task) Action Step Duration 2.Emergency 2.1 Maintain emergency 1 Oversight of the emergency response. 41 sec Direction and direction and control of the Control event response

2. Initiate any emergency actions 48 sec NOTE.

The TMS Is for overlap period. The total time considered for this task should be concurrent with Task 1. END OF EMERGEl'(CY

                                                        ~          .      DIRECTION  ,

AND'CONr.ROL- TASK-

1 min 29 sec

                                                                                '  -   .*,* ,,_TOTAL. TASKD,URATION, COOPER                                                                                                                                                        Page 75

COOPER ON-SHIFT STAFFING ANALYIS REPORT Explain why it Is acceptable or not acceptable for the Shift Manager to perfonn the task of notifying the ERO: The process of activating the ERO takes 48 seconds to perform the actions required followed by 41 seconds of carrying the pager to verify that it activated. Activating the ERO is covered by 2.1-2 to initiate any emergency actions. Per the Shift Manager, more time would be expended directing someone to activate the ERO with code 222 than it would to just do 1t. This task 1s very short 1n duration and the shrft manager is able to stay apprised of plant cond1t1ons while doing 1t. The Shrft Manager used proper place-keeping techniques while performing this task. NAME POSITION DATE Task Performer Assistant Operations Ron Shaw Nov. 3, 2012 Manager

"\.

Evaluator Emergency Preparedness Tim Rients Nov. 3, 2012 Coordinator COOPER Page 76

COOPER ON-SHIFT STAFFING ANALYIS REPORT J. OVERLAP OF TASKS ACTIVITIES OR OTHER CONFLICTS IDENTIFIED A. Overlap Requiring Compensatory Measures None X. REFERENCES

NEI 10-05, Rev 0, Assessment of On-Shift Emergency Response Organization Staffing and Capabilities
NSIR DPR-ISG-01, Interim Staff Guidance - Emergency Planning for Nuclear Power Plants
NUREG-0654, Criteria for PreparatJon and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants.
CNS Emergency Plan, Rev 39 XI. ORIGINAL STAFFING ANAL VIS TEAM
Fred Guynn, Entergy ECH Sr Project Manager, EP
Myra Jones, CMCG Contractor
Dave Werner, CNS Training
John Teten, CNS Chemistry
Kip Reeves, CNS Emergency Planning
Ron Shaw, CNS Operations
Tim Rients, CNS Emergency Planning
Mark Uhri, CNS Chemistry
Greg Brewer, CNS Training COOPER Page77}}

NLS2021001, Emergency Plan, Revision 76, On-Shift Staffing Analysis, Revision 2, and Emergency Plan Implementing Procedures

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2.1 BACKGROUND

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Response

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