NL-15-0820, Vogtle, Units 1 and 2 - Emergency Plan, Revision 64

Vogtle, Units 1 and 2 - Emergency Plan, Revision 64

ML15138A227
Person / Time
Site:	Vogtle
Issue date:	03/31/2015
From:	Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15138A238	List: ML15138A222 ML15138A223 ML15138A225 ML15138A227 ML15138A229 NL-15-0820, Farley, Units 1 and 2 - Emergency Plan, Revision 64, Part 1 of 3 NL-15-0820, Farley, Units 1 and 2 - Emergency Plan, Revision 64, Part 2 of 3 NL-15-0820, Farley, Units 1 and 2 - Emergency Plan, Revision 64, Part 3 of 3 NL-15-0820, Hatch, Units 1 and 2 - Emergency Plan, Revision 37 NL-15-0820, Revision to Emergency Plans NL-15-0820, Vogtle, Units 1 and 2 - Emergency Plan, Revision 64
References
NL-15-0820
Download: ML15138A227 (298)

Text

REV 64 3/15

VOGTLE ELECTRIC GENERATING PLANT UNIT 1 AND UNIT 2 EMERGENCY PLAN

JUSTIFICATIONS FOR CHANGES - REVISION 64 Change Justification 6-1; Added "ETE population reviews will be conducted annually." Added to reflect the process for annual population reviews in accordance with 10 CFR 50 Appendix E. 9-2; Deleted "91304-C, Estimating Offsite Dose" Dose Assessment was standardized throughout the Fleet with NMP-EP-104, Dose Assessment. Site specific procedures were deleted. 9-3; Deleted "NMP-EP-303, Drill and Exercise Standards" The listed procedure was deleted because they had inaccurately been added to the list of Emergency Plan Implementing Procedures. D-31, RG1; Editorial. Deleted misplaced "-" from RE-12444E reading The negative sign (-) had inaccurately been added to the

Emergency Plan. I-3, I.3; replaced "91304-C with NMP-EP-104, Dose Assessment" Dose Assessment was standardized throughout the Fleet with NMP-EP-104, Dose Assessment. Site specific procedures were deleted. Figure J-2; Deleted Follow Up Protective Action

Recommendations Flowchart and replaced with current Flowchart to add statement to consider use of KI with each PAR This addition was made to ensure that every PAR given reflects the consideration for the use of KI. This statement was already reflected in initial PAR flowchart. Index of Effective pages; Updated Revision Number Updated to reflect current version number

i REV 51 10/08 TABLE OF CONTENTS Section Title Page PREFACE xix

A ASSIGNMENT OF RESPONSIBILITIES A-1

A.1 Vogtle Electric Generating Plant A-1

A.2 State of Georgia A-2

A.3 Burke County, Georgia A-8

A.4 State of South Carolina A-9

A.5 Aiken, Barnwell, and Allendale Counties, South Carolina A-14

A.6 Department of Energy - Savannah River Site A-16

A.7 Medical Support A-16

A.8 Private Sector Organizations A-16

A.9 Federal Government Support A-17

A.10 Concept of Operations A-17

B VEGP EMERGENCY ORGANIZATION B-1

B.1 Normal Plant Organization B-1

B.2 Emergency Organization B-1

C EMERGENCY RESPONSE SUPPORT AND RESOURCES C-1

C.1 State and Local Governmental Support C-1

C.2 Federal Governmental Support C-1

C.3 VEGP Support C-3

C.4 Other Support C-4

D EMERGENCY CLASSIFICATION SYSTEM D-1

D.1 Summary of Emergency Classification Logic D-1

D.2 Emergency Class Description and Response D-1

D.3 Classification Process D-6

D.4 Definitions D-15 ii REV 43 5/06 TABLE OF CONTENTS (Continued)

Section Title Page E NOTIFICATION METHODS AND PROCEDURES E-1

E.1 Notification of Personnel E-1

E.2 Notification of State and Local Response Personnel E-2

E.3 Notification of Federal Agencies E-3

E.4 Notification of the Public E-3

E.5 Followup Emergency Messages E-4

E.6 Verification of Notification Messages E-4

F EMERGENCY COMMUNICATIONS F-1

F.1 Communications With the State of Georgia and Burke County F-1

F.2 Communications With the State of South Carolina and Aiken, Barnwell, and Allendale Counties F-2

F.3 Communications With the Savannah River Site F-2

F.4 Communications With Nuclear Regulatory Commission and Other Federal Agencies F-3

F.5 Communications Among VEGP Emergency Response Facilities F-3

F.6 Medical Support Facility Communications F-6

F.7 Alerting Emergency Response Personnel F-7

F.8 Communications System Tests F-7

F.9 VEGP Radiological Monitoring Teams F-7

G PUBLIC EDUCATION AND INFORMATION G-1

G.1 Information for Transients G-2

iii REV 57 06/12 TABLE OF CONTENTS (Continued)

Section Title Page G.2 Joint Information Center Operations G-3

G.3 Offsite Agency Coordination G-4

G.4 Rumor Control G-4

G.5 Media Education G-4

H EMERGENCY FACILITIES AND EQUIPMENT H-1

H.1 Emergency Facilities H-1

H.2 News Center Facilities H-8

H.3 Activation and Staffing of Emergency Facilities H-8

H.4 Plant Monitoring and Data Handling Systems H-9

H.5 Out-of-Plant Monitoring H-17

H.6 Emergency Kits H-18

I ACCIDENT ASSESSMENT I-1

I.1 Plant Parameters I-1

I.2 Radiological Monitors I-1

I.3 Determination of Release Rate I-3

I.4 Dose Assessment System I-3

I.5 Field Monitoring I-5

I.6 Environmental Samples I-8

I.7 Use of Field Monitoring Data I-9

J PROTECTIVE RESPONSE J-1

J.1 Protective Response for Onsite Personnel J-1

J.2 Protective Response for the Public J-4

K RADIOLOGICAL EXPOSURE CONTROL K-1

K.1 Emergency Exposure Guidelines K-1

iv REV 21 10/94 TABLE OF CONTENTS (Continued)

Section Title Page K.2 Onsite Radiation Protection Program K-2

K.3 Decontamination K-3

K.4 Onsite Radiological Contamination Control K-4

L MEDICAL AND PUBLIC HEALTH SUPPORT L-1

L.1 Onsite Capability L-1

L.2 Medical Transportation L-1

L.3 Offsite Support Services L-2

L.4 Training of Medical Support Personnel L-4

M RECOVERY AND REENTRY PLANNING AND POST-ACCIDENT M-1 OPERATIONS

M.1 Commencement of Recovery Phase M-1

M.2 Recovery Organization M-2

M.3 Reentry Planning M-4

M.4 Exposure Monitoring M-4

N EXERCISES AND DRILLS N-1

N.1 Exercises N-1

N.2 Drills N-1

N.3 Scenarios N-4

N.4 Evaluations and Corrective Actions N-5

O RADIOLOGICAL EMERGENCY RESPONSE TRAINING O-1

O.1 Training O-1

O.2 Qualification O-2

O.3 Summary O-2

P RESPONSIBILITY FOR THE PLANNING EFFORT P-1

v REV 46 7/07 TABLE OF CONTENTS (Continued)

Appendix Title 1 GLOSSARY

2 LETTERS OF AGREEMENT AND MEMORANDUMS OF UNDERSTANDING

3 MEANS FOR PROVIDING PROMPT ALERTING AND NOTIFICATION OF THE PUBLIC

4 EMERGENCY EQUIPMENT LIST

5 MEMORANDUM OF AGREEMENT WITH DEPARTMENT OF ENERGY -

SAVANNAH RIVER

6 EVACUATION TIME ESTIMATES FOR THE VOGTLE ELECTRIC GENERATING PLANT PLUME EXPOSURE PATHWAY EMERGENCY

PLANNING ZONE

7 EMERGENCY OPERATIONS FACILITY

8 EMERGENCY COMMUNICATIONS PLAN SOUTHERN NUCLEAR OPERATING COMPANY

9 INDEX OF VEGP ADMINISTRATIVE, NUCLEAR MANAGEMENT AND EMERGENCY PLAN IMPLEMENTING PROCEDURES CROSS-REFERENCED

TO SECTIONS OF THE VEGP EMERGENCY PLAN

10 FSAR-POSTULATED TRANSIENTS

11 LETTER FROM THE STATE OF GEORGIA

vi REV 27 02/98 LIST OF TABLES Table Title A-1 Responsible Individuals of Primary Response Organizations

B-1 Minimum Staffing for Power Operation

B-2 Emergency Organization Assignments

C-1 Airport Characteristics

C-2 State and County Emergency Operation Centers (EOC)

D-1 Hazards and Other Conditions Affecting Plant Safety for All Modes

D-2 System Malfunction Emergency Action Levels

D-3 Abnormal Rad Levels/Radiological Effluent

D-4 Fission Product Barriers Status

E-1 Initial Notification System (Normal Working Hours)

E-2 Initial Notification System (Backshift Hours)

F-1 Emergency Response Communications Summary

H-1 Integrated Plant Computer Inputs

vii REV 63 10/14 LIST OF TABLES (Continued)

Table Title J-1 Use of Protective Equipment and Supplies J-4 Evacuation Time Estimate Summary

J-5 Reception Centers/Shelters for Counties Within the Plume Exposure Pathway EPZ of VEGP

K-1 Emergency Exposure Limits Per Person Per Response

O-1 Training Course Descriptions

O-2 Training Requirements for VEGP ERO Personnel

viii REV 38 05/05 LIST OF FIGURES Figure Title i General Vicinity Map

ii Vogtle Electric Generating Plant Site Plan

iii Vogtle Electric Generating Plant 10-Mile EPZ (Plume Exposure Pathway)

iv Vogtle Electric Generating Plant & Savannah River Site 50-Mile EPZ (Ingestion Pathway)

v 1987 Permanent Population Within the Vogtle Electric Generating Plant Plume Exposure EP2

vi Transient and Special Facility Population (Employees, Visitors to Recreational Areas and Schools) Within the Vogtle Electric Generating Plant Plume Exposure EP2

A-1 Formal Interfaces Among Emergency Organizations

A-2 State of Georgia Operating in a Declared Radiological Emergency

A-3 Operational Relationships Among Burke County Response Organizations

A-4 State of South Carolina Operating in a Declared Radiological Emergency

A-5 Aiken County, S.C., Emergency Response Organization

A-6 Barnwell County, S.C., Emergency Response Organization

A-7 Allendale County, S.C., Emergency Response Organization

B-1 VEGP Organization Chart

B-2 Response Organization for Alert

B-3 Site Area or General Emergency Response Organization

ix REV 63 10/14 LIST OF FIGURES (Continued)

Figure Title E-1 Example of Initial Emergency Message For State and Local Response Agencies

E-2 Example of NRC Operations Center Event Notification Worksheet

J-1 Initial Protective Action Recommendations J-2 Follow Up Protective Action Recommendations

H-1 TSC Layout Plan

H-2 OSC Layout Plan

M-1 Recovery Organization

P-1 Emergency Preparedness Organizations VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

x REV 64 3/15 Table of Contents (i) 51 ii 43 iii 57 iv 21 v 46 vi 27 vii 63 viii 38 ix 63 x 64 xi 64 xii 60 xiii 64 xiv 64 xv 64 xvi 60 xvii 60 xviii 64 Preface xix 39 xx 41 Figure i 5 Figure ii 27 Figure iii 14 Figure iv 23 Figure v NONE Figure vi 12 A-1 56 A-2 30 A-3 23 A-4 30 A-5 30 A-6 30 A-7 23 A-8 18 A-9 61 A-10 61 A-11 35 A-12 35 A-13 31 A-14 23 A-15 35 A-16 61 A-17 56 A-18 23 A-19 35 A-20 38 A-21 24 Table A-1 (1 of 2) 35 Table A-1 (2 of 2) 14 Figure A-1 35 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV. xi REV 64 3/15 Figure A-2 22 Figure A-3 2 Figure A-4 35 Figure A-5 35 Figure A-6 35 Figure A-7 35 B-1 61 B-2 61 B-3 57 B-4 57 B-5 37 B-6 37 B-7 61 B-8 38 B-9 56 B-10 19 Table B-1 (1 of 2)60 Table B-1 (2 of 2)60 Table B-2 (1 of 2)60 Table B-2 (2 of 2)43 Figure B-1 50 Figure B-2 43 Figure B-3 43

C-1 61 C-2 61 C-3 61 C-4 61 C-5 61 Table C-1 (1 of 2)2 Table C-1 (2 of 2)28 Table C-2 56 Section D (pp D-1 to D-5)48

D-6 56 D-7 48 D-8 48 D-9 48 D-10 48 D-11 48 D-12 48 D-13 48 D-14 to D-16 57

D-17 62 D-18 57 D-19 to D-21 57

D-22 62 D-23 to D-25 62 D-26 to D-28 57

D-29 62 D-30 57 D-31 64 D-32 57 D-33 48 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV. xii REV 62 10/14 D-34 56 D-35 56 D-36 48 D-37 48 D-38 48 D-39 48 D-40 48 D-41 48 D-42 48 D-43 62 D-44 48 D-45 62 D-46 to D-49 48

D-50 62 D-51 48 D-52 62 D-53 48 D-54 62 D-55 to D-56 48 D-57 to D-58 60

D-59 48 D-60 62 D-61 48 D-62 62 D-63 62 D-64 53 D-65 48 D-66 62 D-67 62 D-68 to D-71 48

D-72 56 D-73 48 D-74 to D-75 55 D-76 to D-83 48 D-84 to D-85 55 D-86 to D-89 48 D-90 to D-92 55 D-93 48 D-93a 60 D-94 to D-97 48

D-98 62 D-99 48 D-100 48 D-101 48 D-102 to D-116 48

E-1 61 E-2 60 E-3 56 E-4 56 Table E-1 (1 of 2)35 Table E-1 (2 of 2)25 Table E-2 (1 of 2)33 Table E-2 (2 of 2)33 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

xiii REV 64 3/15 Figure E-1 43 Figure E-2 (1 of 3) -DELETED 21 Figure E-2 (1 of 2) 35 Figure E-2 (2 of 2) 35 F-1 56 F-2 43 F-3 30 F-4 28 F-5 57 F-6 60 F-7 43 Table F-1 60 G-1 24 G-2 41 G-3 60 G-4 60 H-1 51 H-2 23 H-3 57 H-4 30 H-5 61 H-6 38 H-7 38 H-8 57 H-9 38 H-10 35 H-11 35 H-12 56 H-13 56 H-14 38 H-15 38 H-16 22 H-17 34 H-18 57 Table H-1 34 Figure H-1 34 Figure H-2 57 I-1 56 I-2 57 I-3 64 I-4 25 I-5 38 I-6 61 I-7 38 I-8 57 I-9 57 I-10 21 J-1 35 J-2 39 J-3 61 J-4 38 J-5 63 J-6 11 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

xiv REV 64 3/15 Table J-1 22 Table J-2 63 Table J-3 63 Table J-4 61 Table J-5 (1 of 2)39

Table J-5 (2 of 2)

Figure J-1

Figure J-2 41 63 64 K-1 57 K-2 21 K-3 57 K-4 23 Table K-1 23 Table K-2(DELETED) 23 L-1 60 L-2 56 L-3 56 L-4 19 M-1 37 M-2 19 M-3 60 M-4 60 Figure M-1 60 N-1 61 N-2 61 N-3 61 N-4 61 N-5 61 N-6 N-7 61 61 O-1 35 O-2 50 Table O-1 (1 of 3) 57 Table O-1 (2 of 3) 57 Table O-1 (3 of 3) 39 Table O-2 (1 of 2) 56 Table O-2 (2 of 2) 46 P-1 62 P-2 62 P-3 61 Figure P-1 62 Appendix 1 0 1-1 39 1-2 39 1-3 49 1-4 56 1-5 22 Appendix 2 46 2-1 62 2-2 46 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

xv REV 64 3/15 Appendix 3 0 3-1 61 3-2 19 3-3 5 3-4 56 3-5 52 3-6 31 3-7 31 3-8 38 3-9 38 3-10 38 Figure 3-1 7 Appendix 4 0

4-1 60 4-2 49 4-3 25 4-4 60 4-5 60 4-6 60 4-7 60 Appendix 5 0 Department of Energy -Savannah River 04/12/99

Site Memorandum of Agreement (4 pages)5 Appendix 6 59

6-1 64 6-2 6-3 61 61 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

xvi REV 60 07/13 Appendix 7 Emergency Operations Facility 38 Page 1 38 Page 2 38 Page 3 38 Page 4 60 Page 5 38 Page 6 38 Page 7 43 Page 8 38 Page 9 38 Page 10 38 Page 11 60 Page 12 38 Page 13 38 Page 14 38 Page 15 43 Page 16 38 Page 17 38 Page 18 38 Page 19 38 Page 20 38 Page 21 38 Page 22 38 Page 23 38 Page 24 38 Page 25 38

VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

xvii REV 60 07/13 Page 26 38 Page 27 49 Page 28 49 A ttachment B (Mutual Assistance Agreements) DELETED 21

1) Nuclear Power Plant Assistance A gree. (16 pgs.)DELETED 21 2) Electric Utilities Assist A greement (15 pages) DELETED 21 Appendix 8 (Emergency Commun. Plan) 57 Table of Contents (i)57

ii 57 iii 57 1 57 2 57 3 57 4 57 5 57 6 57 7 57 8 57 9 57 10 57 11 57 12 60 13 60 14 60 15 57 16 60 17 60 18 57 19 57 20 57 21 57 22 57 23 57 24 57 25(TABLE 1) 57

26 60 27 60 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV.

xviii REV 64 3/15 28 57 29 60 30 60 Appendix 9 57 9-1 57 9-2 64 9-3 64 Appendix 10 1 Sheet 1 of 2 21 Sheet 2 of 2 21 Appendix 11 6 GEMA Letter dated 9-7-84 0

xix REV 39 11/04 PREFACE The Vogtle Electric Generating Plant (VEGP) is a two-unit pressurized water reactor operated by Southern Nuclear

Operating Company (SNC). The plant is on a 3169-acre site

located in the eastern portion of Burke County, Georgia, approximately 23 river mi upstream from the intersection of the

Savannah River with U.S. Highway 301, as shown on figure i.

Figure ii shows the site and the locations of the buildings on

the site. The locations of the VEGP emergency facilities are shown on figure ii.

This Emergency Plan is applicable to VEGP, Units 1 and 2, and

to its environs as specified by the emergency planning zones (EPZs): a plume exposure pathway EPZ, which nominally consists

of the area within approximately 10 mi of the plant, and an

ingestion exposure pathway EPZ, which extends to 50 mi. These

distances are reckoned from a point midway between the centers

of the Unit 1 and Unit 2 containment buildings for the 10 mile

EPZ map. The two EPZs are shown in figures iii and iv.

Because of the location of VEGP, the emergency planning and/or

protective action responsibilities at the state level involves

two states, Georgia and South Carolina. Both Georgia and South

Carolina, as well as the counties (Burke County in Georgia and

Aiken, Barnwell, and Allendale Counties in South Carolina)

within the plume exposure pathway EPZ, have prepared plans for

a response to an emergency at VEGP. These plans describe their

respective responsibilities, authorities, capabilities, and

emergency functions. The major portion of the plume exposure

pathway EPZ in South Carolina is within the Department of

Energy's Savannah River Site (SRS). The Department of Energy, Savannah River Operations Office (DOE-SR), pursuant to a

memorandum of agreement between Georgia Power Company (GPC), as

assigned to SNC, and DOE-SR provided in appendix 5, will be

responsible for all emergency response actions on the SRS

whenever an emergency occurs at VEGP.

Small areas of Georgia within a 10-mi radius of the plant have

been excluded from the detailed state and local planning. The

area in Richmond County has been excluded because it is a

low-lying swampland with no residences, no access routes, and

limited recreational use. A letter documenting this provision is

included in appendix 11.

Within the plume exposure pathway EPZ, Burke County in the

State of Georgia has the largest resident population. However, even here the population is small and dispersed. The areas in

South Carolina that are not federally owned or controlled are

along the Savannah River lowlands in Aiken, Allendale, and

Barnwell Counties. The segment in Aiken County, approximately

8 to 10 mi NNW from VEGP, is part of the Cowden Plantation

xx REV 41 12/05 and has no resident population. The segments to the ESE, Barnwell and Allendale Counties (approximately 9 to 10 mi from

VEGP), are largely comprised of portions of the Creek Plantation, a horse farm. Within the South Carolina portion of the plume

exposure pathway EPZ, the only housing occurs within the Creek

Plantation in Barnwell County, where there are only a limited

number of permanent residences. Figure v presents the permanent

population within the plume exposure pathway EPZ.

The transient population within the plume exposure pathway EPZ

includes persons in the work force at the VEGP and at

recreational areas, mainly hunters and fishermen (see figure vi).

This transient population is generally along the Savannah River;

around the Cowden Plantation and Gray's Landing in Aiken County;

around Creek Plantation in Barnwell and Allendale Counties; and

at St. Mary's Baptist Church in Barnwell County.

The EPZ for ingestion exposure includes an area within 50 mi of

VEGP. Planning for the ingestion exposure pathway is a

responsibility of the States of Georgia and South Carolina.

More information about the ingestion exposure pathway EPZ can

be obtained from the States' Radiological Emergency Plans.

The order of the presentation provided herein follows that of

the 16 standards delineated in 10 CFR 50, Section 50.47 (b).

Appropriate criteria from NUREG-0654, Revision 1, Criteria for

Preparation and Evaluation of Radiological Emergency Response

Plans and Preparedness in Support of Nuclear Power Plants, are

addressed approximately in the sequence presented in that

document.

Although this Plan is designed to stand on its own, two

appendices expand on matters mentioned here: the Emergency Operations Facility (appendix 7) and the Emergency Communication Plan (appendix 8). It is to be recognized that

this is only a plan and not a prescriptive document. Each

incident is a unique event; therefore this Plan is designed to

incorporate the flexibility to tailor the response and meet the

emergency.

This Plan is supported by a set of implementing procedures. A

listing of these procedures is included as appendix 9.

FIGURE ii Vogtle Electric Generating Plant Site Plan FIGURE iii Vogtle Electric Generating Plant 10-Mile EPZ (Plume Exposure Pathway)

FIGURE iv Vogtle Electric Generating Plant & Savannah River Site 50-Mile EPZ (Ingestion Pathway)

These figures are available in hard copy only.

See Christopher E. Boone, Emergency Planning Coordinator (205) 992-6635

A-1 REV 56 09/10 A. ASSIGNMENT OF RESPONSIBILITIES In the event of a situation at the Vogtle Electric Generating Plant (VEGP) which requires activation of the emergency

response organizations, various State, local, Federal, and

private sector organizations may be required to contribute to

the emergency response. This section describes the

responsibilities of these organizations. Table A-1 lists

primary response organizations and the emergency title of the

individual in charge.

A.1 VOGTLE ELECTRIC GENERATING PLANT Vogtle Electric Generating Plant (VEGP) accepts the responsibility of maintaining an effective Emergency Plan and

maintaining proper preparedness through the maintenance of

formal procedures for implementing the Plan as identified in

appendix 9, the training of personnel in accordance with

section O, the maintenance of necessary equipment, and

continuing relationships with various governmental agencies

and private organizations as identified in this section. The following tasks are part of VEGP's responsibility:

1. Recognize and declare the existence of an emergency condition.

2. Classify the event in accordance with the methodology described in section D of this Plan.

3. Notify appropriate VEGP personnel and offsite authorities.

4. Take corrective actions to mitigate the severity of the accident.

5. Request additional support as deemed necessary.

6. Establish and maintain effective communications within VEGP and with offsite response groups as described in section F.

7. Continuously assess the status of the accident and periodically communicate the status information to

the appropriate response groups. This includes the

collection and evaluation of onsite and offsite

radiological monitoring data.

8. Take protective measures onsite and recommend protective measures to offsite authorities.

9. Monitor and control radiation exposures of all personnel responding to the emergency and under the

direction of VEGP.

A-2 REV 30 07/00 10. Provide emergency information to the public through periodic press briefings in conjunction with State and local officials.

The VEGP emergency response is carried out under the control of

the emergency director. The onsite organization to support these

activities is described in section B of this Plan. A.2 STATE OF GEORGIA Georgia has developed a Radiological Emergency Response Plan on a statewide basis as an integral part of the Georgia Emergency

Operations Plan (GEOP). The GEOP is an emergency operations plan for all natural disasters, accidents, and incidents, including

radiological emergencies at fixed nuclear facilities. It is a

plan of action developed for use by local and State government

officials in preparing for, responding to, and dealing with

situations throughout the State.

In accordance with the Governor's Executive Order dated July, 1992, the Georgia Department of Natural Resources (DNR) has the lead agency responsibility for responding to all peacetime radiological emergency situations throughout Georgia. Under the

Emergency Support Function (ESF) 10, Hazardous Materials of the GEOP, which was developed pursuant to the Governor's Executive Order, the DNR radiological emergency response team assesses the radiological conditions at the site of an incident and determines

whether a state of emergency exists. Upon being advised that a

radiological emergency exists, the Governor declares an emergency

condition, which then activates the Georgia Emergency Management

Agency (GEMA) authorities to deal with the situation. Under the

statutory authority granted to GEMA, the preestablished plans and

procedures of eight State agencies and local government

organizations are automatically activated and coordinated by the

State Operations Center (SOC) in Atlanta or the GEMA Forward Emergency Operations Center (FEOC) in Waynesboro. In the event

of a radiological emergency, GEMA has broad legal authority to

take whatever actions are deemed necessary to protect the health

and safety of Georgia citizens. This authority includes, but is

not limited to, evacuation of people from private property, control of public and private transportation corridors, and

utilization of all public facilities in support of efforts to

protect life and property.

The fundamental legislation providing the basis for emergency response by civil authorities is the Georgia Emergency Management

Act of 1981, as amended. This Act in part creates a State

Emergency Management Agency; authorizes the creation of local

organizations for emergency management; confers upon the Governor

and upon the executive heads of governing bodies of the State

certain emergency powers; and provides the rendering of mutual

aid among the political subdivisions of the State, and with other

states, and with the Federal Government.

A-3 REV 23 4/96 Other documents providing bases for emergency response include:

1. Governor's Executive Order, August 25, 1981:

Recognizes the Georgia Emergency Management Act of

1981, which redesignates the State Civil Defense

Agency as the Georgia Emergency Management Agency.

2. Georgia Emergency Operations Plan: Contains the rules and regulations for operations, should an

emergency or disaster occur in the State. The Plan

is binding on all local governments authorized or

directed to conduct emergency management operations

and on all State departments or agencies.

3. Radiation Control Act, Georgia Code Annex 88-1301 et seq: Delegates emergency powers during radiation

emergencies to the Department of Natural Resources, Division of Environmental Protection.

4. Georgia Water Quality Control Act of 1974, as amended, Act No. 870.

5. Georgia Air Quality Control Act of 1978, as amended, Act No. 794.

6. Georgia Transportation of Hazardous Materials Act of 1979, Act No. 487.

The duties and responsibilities of the principal and support

agencies of the State of Georgia are summarized below. A

detailed discussion of the State's response is contained in

the Georgia Radiological Emergency Response Plan.

A.2.1 PRINCIPAL AGENCIES OF THE STATE OF GEORGIA The following State agencies are assigned lead responsibility for radiation emergencies and for overall State preparedness, respectively:

1. GEORGIA EMERGENCY MANAGEMENT AGENCY

a. GEMA is responsible for general State emergency planning and exercises overall direction and

control of emergency or disaster operations as

assigned by Executive order.

b. The director of emergency management as the State disaster coordinator coordinates emergency

activities of DNR with overall State response

efforts.

A-4 REV 30 07/00

c. During an emergency, the State of Georgia, acting through GEMA, and the State of South Carolina, acting through the South Carolina Emergency

Preparedness Division, will coordinate their

emergency response efforts including such matters

as radiological dose assessment, protective

action decisions, and activation of the prompt

notification system for alerting the public.

d. On behalf of the Governor, activate all or portions of the GEOP to provide the necessary overall coordinated response.

e. Provide communications for the State Operations Center (SOC), as required, through 24-hour radio network, commercial telephone, National Warning

System (NAWAS), or other communications systems.

Communication links will be established in

accordance with existing procedures with the SOC in Atlanta as well as with additional State and

local emergency response personnel within the

plume exposure pathway and ingestion EPZs. These

functions will initially be handled from the SOC and will then be transferred to the Forward

Emergency Operations Center (FEOC) once

activated.

f. Establish the FEOC if necessary. The FEOC will be established at the Burke County Emergency

Operations Center (EOC) Building, on the corner

of Georgia Highway 24 and Perimeter Road in

Waynesboro. Upon the decision of the GEMA

director to activate the FEOC, the GEMA Mobile

Communications Vehicle (MCV)may be dispatched to Burke County and arrive within driving time plus

30 min for mobilization. When dispatched, it may

be used to transport certain State responders.

The GEMA MCV will be located adjacent to the Burke County EOC and will provide support

services to State agencies at the FEOC. The

Burke County EOC has the necessary electrical

hook-ups to provide long term electrical services

to the vehicle. In addition to providing backup

for the communications described in the preceding

paragraph, the MCV provides backup communications for State agencies other than GEMA (Department of

Transportation, the Georgia State Patrol, the

Georgia Department of Natural Resources - Law

Enforcement Section, and the Georgia Forestry

Commission). In addition, the MCV has an on-board weather station that can provide wind speed and direction, barometric pressure and relative humidity. It also has micro down-link capabilities and a video camera system.

A-5 REV 30 07/00

g. Maintain liaison with DNR radiation emergency coordinator (REC).

h. Activate public emergency warning and/or evacuation procedures, as needed, pursuant to the GEOP. i. Assist radiological monitoring and provide instrumentation.

j. Provide radiological monitoring training assistance.

k. Assist in area security and control; request National Guard assistance if needed.

l. Provide for the coordination of land and air transportation for emergency personnel as

requested.

m. Coordinate public information releases in cooperation with State and local agencies.

n. Communication links will be established with the state of South Carolina Emergency Preparedness

Division and DOE-SRS to ensure efficient

coordination of emergency response activities.

2. DEPARTMENT OF NATURAL RESOURCES

a. DNR is assigned primary responsibility by Executive Order for implementation and

administration of the State radiological

emergency response function. DNR will: (1) Dispatch radiation emergency teams as needed. (2) Perform radiation survey and monitoring.

(3) Provide radiation safety training.

(4) Direct recovery/reentry operation and provide health physics control of contaminated areas.

b. A radiation emergency coordinator (REC) in the Environmental Protection Division (EPD) interacts with appropriate State, local, and Federal

agencies and private organizations to direct all

necessary radiation control actions. The REC is

on call 24 h and will be notified by the GEMA

duty officer.

c. In situations beyond local government control, DNR provides program assistance in the

application of available personnel, equipment, and technical expertise as required.

d. DNR requests State support agency(s) and Federal assistance pursuant to the GEOP as required.

A-6 REV 30 07/00

e. DNR will provide for health physics escort of media and other personnel within the plume exposure pathway emergency planning zone (EPZ) as conditions allow if access controls have been

established. A.2.2 State Support Agencies in Georgia

The following State agencies are prepared to provide related

support of this function as indicated pursuant to the GEOP:

1. DEPARTMENT OF HUMAN RESOURCES Coordinate emergency health and social assistance pursuant to the GEOP.

2. DEPARTMENT OF PUBLIC SAFETY

a. As applicable, assume control over the onsite situation until the arrival of radiation safety personnel.

b. Maintain liaison with DNR REC.

c. Provide communication linkage as required.

d. Provide land or air transportation or escort as available for radiation safety personnel, other necessary personnel, or equipment.

e. Assist in radiological monitoring as required.

f. Provide law enforcement assistance for area security or recovery of lost or stolen radioactive material.

g. Coordinate with DNR law enforcement and local police. h. Assist in public warning or evacuation as required.

3. DEPARTMENT OF AGRICULTURE

a. Collect samples of food products, livestock, produce, and dairy products, as necessary.

b. Restrict the sale, production, distribution, and warehousing of livestock, produce, dairy, and

processed food products contaminated beyond safe

consumption.

c. Assist in disposal of contaminated products.

A-7 REV 23 4/96

d. Coordinate these activities with United States Department of Agriculture (USDA) personnel.

e. Maintain liaison with DNR REC for assessing degree of contamination.

4. DEPARTMENT OF TRANSPORTATION

a. Assist in traffic control and routing, accident assessment, and recovery operations in transportation incidents.

b. As requested, provide land, air, or water transportation for radiation safety personnel, other necessary personnel, or equipment.

c. Provide communications linkage as required.

d. Assist State Patrol and DNR law enforcement in security and radioactive material escort as requested.

e. Provide heavy equipment and personnel as required.

5. FORESTRY COMMISSION

a. Provide land or air transportation as requested for radiation safety personnel, other necessary

personnel, or equipment.

b. Provide personnel and heavy equipment as required to assist in recovery operations.

c. Provide communication linkage as necessary.

6. DEPARTMENT OF ADMINISTRATIVE SERVICES

a. Provide for expeditious approval and purchase of equipment and supplies essential to emergency

operations.

b. Provide land transportation vehicles for emergency personnel.

c. Provide emergency communications equipment and repair.

A-8 REV 18 5/93 A.3 BURKE COUNTY, GEORGIA All the area within the plume exposure pathway EPZ in the State of Georgia falls within Burke County. The

responsibility for overall radiological emergency response

planning rests with the chairman, Burke County Board of

Commissioners. It is the chairman's responsibility to

initiate actions and provide direction and control at a level

consistent with the specific incident. Agencies within Burke

County which have a primary role in radiological emergency

planning and response include the Emergency Management Agency

and the Sheriff's Department. A detailed discussion of the

Burke County response is contained in Annex D to the Georgia

Radiological Emergency Response Plan.

The chairman, Burke County Board of Commissioners, or his

predesignated alternate (the vice chairman, any of three

other commissioners, the county administrator, or the EMA

director), may declare an emergency and implement offsite

protective actions. However, upon the declaration of an

emergency situation or the decision to implement protective

actions the emergency management agency director is

authorized to represent the chairman, Burke County Board of

Commissioners.

A.3.1 Emergency Management Agency

Principal activities include the following:

1. Receive notification from VEGP and GEMA.

2. Maintain communications with VEGP on emergency situation status.

3. Provide information to other Burke County response elements, to the media, and to the public.

4. Activate the public notification system, if required.

6. Coordinate Burke County emergency response activities.

7. Activate and direct operations at the designated reception and care facility.

8. Implement protective action recommendations as requested by GEMA or the chairman of the Burke County Board of Commissioners or his predesignated

designee.

A-9 REV 61 06/14 A.3.2 Sheriff's Department

Principal activities include the following:

1. Provide backup public notification.

2. Control access to the plume exposure pathway EPZ.

3. Provide traffic control and law enforcement measures in the event of an evacuation.

A.3.3 Fire Support VEGP has established an agreement with the Burke County EMA to provide, upon request, offsite fire support to the VEGP Fire Brigade. Support provided includes, but is not limited to, firefighters and firefighting equipment. Request for fire support will be made by the control room or site security to the Burke County 911 center, Burke County EOC, or the Incident Command Post, as applicable, based on the nature and timing of the event. A copy of this agreement is maintained in the SNC document management system and is included by reference in Appendix 2.

A.3.4 Others Other county resources, including the Fire Department, Health

Department, and Public Works Department, may be mobilized as

described in Annex D to the Georgia Radiological Emergency

Response Plan.

A.4 STATE OF SOUTH CAROLINA The State of South Carolina has developed the South Carolina

Radiological Emergency Response Plan (SCORERP) which provides

guidance to state and local governments on procedures, organization, and responsibilities for preventing and

mitigating the effects of a nuclear power plant incident or

disaster. The SCORERP provides for the radiological

emergency response in the event of a radiological accident at

a fixed nuclear facility located in the state or in states

contiguous to South Carolina. The SCORERP describes the

South Carolina Radiological Emergency Response Organization, which consists of the Office of the Adjutant General (Emergency Management Division (EMD), the Department of

Health and Environmental Control (Bureau of Solid and

Hazardous Waste, Nuclear Emergency Planning (NEP) Section),

and those state resources available to local government(s)

during a fixed nuclear facility radiological accident. State

radiological emergency response forces will be operational

upon notification of a radiological emergency. The response

functions of involved agencies are described in the SCORERP A-10 REV 61 06/14 and the South Carolina Technical Radiological Emergency Response Plan (SCTRERP). A VEGP Site Specific Radiological

Emergency Response Plan (Part 5 to SCORERP) addresses those

matters that pertain directly to VEGP.

Upon being advised by EMD that a radiological emergency

exists, the Governor or his designated representative

declares an emergency condition. Under the Governor's

direction, the total and combined efforts of the state and

local governments will be utilized to nullify or reduce the

effects of offsite radiological hazards resulting from a

nuclear plant accident. As deemed necessary, the Governor may

transfer the direction, personnel, or functions of state

agencies. All radiological emergency response organizations

are prepared to react on a 24-h basis, and they are capable

of continuous operations for a protracted period. Directors

of state agencies, departments, and commissions are

responsible for ensuring that their agencies' radiological

emergency response responsibilities are accomplished.

Designated county officials are responsible for emergency

response within their jurisdictions.

The legal basis and authority for the emergency response of

the State of South Carolina include:

1. South Carolina Constitution, Article IV, Annex 17, Appendix A

2. South Carolina Code of Laws ANN, 25-1-420 through 25-1-460 3. Regulation 58-1, Local Government Preparedness Standards; and Regulation 58-101, State Emergency

Preparedness Standards, South Carolina Code of

Regulations.

4. The South Carolina Emergency Operations Plan (SCEOP). 5. The South Carolina Operational Radiological Emergency Response Plan (SCORERP).

6. The South Carolina Technical Radiological Emergency Response Plan (SCTRERP).

A-11 REV 35 6/02 The duties and responsibilities of the principal and support agencies of the State of South Carolina are summarized below.

A.4.1 Principal Agencies of the State of South Carolina

The following state agencies are assigned primary

responsibilities for overall state radiological emergency

management and response operations.

1. Office of the Governor

a. Upon being advised that a radiological emergency exists, the Governor declares an emergency

condition which results in the activation of EMD, DHEC, and other state agencies to deal with the

situation, in accordance with their emergency

response plans.

b. The Governor's press secretary or authorized representative will direct and control public

information service activities and serve as the

Governor's official representative/spokesperson

regarding the preparation and release of

emergency information by state government.

2. Office of the Adjutant General Emergency Management Division

a. The EMD is assigned the responsibility for coordinating the emergency planning efforts of all state, county, and municipal agencies in

developing a State emergency plan; conducting a

preparedness program to assure capability of the

government to execute the plan; establishing and

maintaining a State EOC and providing support of

the State emergency staff and work force; and

establishing an effective system for reporting, analyzing, and disseminating emergency

information.

b. The director of EMD is the principal advisor to the Governor for emergency response. He

coordinates with DHEC the recommended protective

measures based on DHEC evaluation and assessment

of the radiological emergency at VEGP.

A-12 REV 35 6/02

3. Department of Health and Environmental Control

a. The Department of Health and Environmental Control (DHEC) maintains a radiological hazard assessment capability and provides technical

support, coordination, and guidance for the State

and local governments. It will conduct and/or

coordinate radiological surveillance and

monitoring in coordination with DOE-SRS and VEGP.

DHEC will obtain and coordinate radiological

assistance resources from the Federal Government, other states, and the nuclear industry as

required.

b. The Bureau of Solid and Hazardous Waste, Nuclear Emergency Planning (NEP) Section within DHEC has

the direct responsibility to provide technical

assistance and resources necessary to evaluate

and assess the consequences of a radiological

incident and to provide protective action

guidance to State and local authorities. NEP is

responsible for radiological surveying and

monitoring of the environment, including the

collection and analyses of samples of soil, air, water, milk, and crops.

c. The Manager, Nuclear Emergency Planning Section is the individual responsible for implementing

the SCTRERP; providing EMD with recommended protective actions as well as recovery and

reentry guidelines. NEP has 24-h accident

response capability with EMD. A.4.2 State Support Agencies in South Carolina

The following state agencies are prepared to provide related

support pursuant to SCEOP and SCORERP.

1. Department of Public Safety (Highway Patrol Division)

a. Regulate and control traffic on the highways of the state.

b. Have its Highway Patrol warning point serve as the backup means of notification from the VEGP.

c. Provide for a 24-h communication system with VEGP, EMD, and DHEC.

d. Support the State Law Enforcement Division in security and other law enforcement activities.

e. Assist in transportation accident assessment and recovery operations.

A-13 REV 31 12/00

f. As requested, provide transportation for radiation safety personnel, other emergency response personnel, or equipment.

g. Provide heavy equipment and personnel as requested.

2. State Law Enforcement Division (SLED)

a. Coordinate law enforcement activities, particularly during evacuation.

b. Assume control of areas suspected of contamination until the arrival of radiation

safety personnel.

c. Assist in search and rescue.

d. Coordinate and assist in securing the SEOC,waterways and evacuated areas.

f. Augment communication system and warning points operation.

3. Forestry Commission

a. Provide land or air transportation as requested for emergency response personnel or equipment.

b. Provide fire control service.

c. Assist in decontamination, in coordination with DHEC. d. Provide radio operators for the SEOC to operate permanently installed Forestry Commission radio equipment.

e. Maintain radio contact with all Forestry Commission elements in affected counties.

4. Department of Natural Resources

a. Coordinate search and rescue operations.

b. Provide air transportation in support of radiological monitoring operations.

c. Augment public warning operations and public information services.

d. Assist in law enforcement.

5. Department of Social Services

a. Coordinate shelter operations which includes registration of evacuees.

A-14 REV 23 04/96

b. Coordinate all emergency welfare services to evacuees such as feeding, clothing, and information.

c. Administer and maintain individual and family assistance program.

d. Augment public information service to evacuees.

6. Clemson University Cooperative Extension Service

a. Maintain updated agricultural data required for radiological assessment in support of DHEC.

b. Assist in locating contaminated livestock, feed, milk, and other farm products for disposal or

decontamination by DHEC.

7. Department of Agriculture

a. Restrict the sale, production, distribution, and warehousing of livestock, produce, dairy and

processed food products contaminated beyond safe

consumption.

b. Assist DHEC in disposal of contaminated products.

d. Coordinate these activities with the United States Department of Agriculture (USDA). A.5 AIKEN, BARNWELL, AND ALLENDALE COUNTIES, SOUTH CAROLINA Most of the plume exposure pathway EPZ within South Carolina falls within the site boundary of the Savannah River Site (SRS). The United States Department of Energy is responsible

for the direction and control of all emergency response

actions on the SRS.

There are limited portions of Aiken, Barnwell, and Allendale

counties which are outside of the SRS but within the plume

exposure pathway EPZ of VEGP. These counties are similarly

organized, with the responsibility for overall radiological

emergency response planning resting with the chairman of the

county council in each case. It is the chairman's

responsibility to initiate actions and provide direction and

control at a level consistent with the specific incident.

A-15 REV 35 6/02 Agencies within these counties which have a primary role in radiological emergency planning and response include the

Emergency (Disaster) Preparedness Agency (EPA) and the

Sheriff's Department. A detailed discussion of the county

response is contained in Annex Q2 to each county's emergency

operations plan.

The chairman of the county council, or his predesignated

alternate (the vice chairman, county administrator, or EPA

director) may declare an emergency within his respective

county and implement offsite protective actions. The

chairman or his designee is available 24 h per day.

The locations of the county EOCs are indicated in Table C-2.

Principal emergency response activities include the

following:

A.5.1 Aiken County Emergency Preparedness Division, Allendale County Emergency Preparedness Agency, Barnwell County Emergency Management Agency.

a. Maintain communications with South Carolina EMD and VEGP on emergency situation status.

b. Provide information to other county response elements.

c. Activate the county EOC.

d. Activate public notification system if required.

e. Coordinate county emergency response activities.

f. Activate and direct operations at the designated reception facility.

g. Implement protective actions as requested by South Carolina EMD or the chairman, county council, or his predesignated designee. A.5.2 Sheriff's Department

a. In Barnwell and Aiken Counties, receive notification from VEGP and South Carolina EMD, as primary warning points. This function is

performed by the Allendale County Central

Dispatch.

b. Provide backup public notification.

c. Provide traffic control and law enforcement in the event of an evacuation.

A-16 REV 61 06/14

d. Coordinate access to the plume exposure pathway EPZ. e. Provide security at EOC.

Other county resources, including Fire Department, Public Works Department, Emergency Medical Services, and Department

of Social Services, may be mobilized as described in the

county emergency operations plan.

A.6 DEPARTMENT OF ENERGY - SAVANNAH RIVER SITE A significant portion of the plume exposure pathway EPZ falls within the site boundary of the Savannah River Site (SRS).

The United States Department of Energy - Savannah River

Operations consists of lands owned or leased by the Federal

government. As such, DOE-SR is responsible for the direction

and control of all emergency response actions on the SRS.

See memorandum of agreement between DOE - Savannah River

Operations Office and Georgia Power Company, as assigned to

SNC, (appendix 5).

A.7 MEDICAL SUPPORT VEGP has established agreements with the Burke County EMA to provide ambulance service for the transportation of injured

personnel, including people who may be radioactively

contaminated, to hospital facilities for treatment.

Agreements with Burke Medical Center in Waynesboro, Georgia, and Doctors Hospital in Augusta, Georgia, have also been

established for treatment of injured and contaminated

individuals. Support provided includes, but is not limited to, Emergency medical services, ambulances, and emergency medical technicians. Request for medical support will be made by the control room or site security to the Burke County EMAs of the Incident Command Post, as applicable, based on the nature and timing of the event. Copies of these agreements are maintained in the SNC document management system and are included in Appendix 2. For radiological medical emergency training, agreements have been established

with Tetra Tech.

A.8 PRIVATE SECTOR ORGANIZATIONS

1. BECHTEL POWER CORPORATION GPC/SNC has established an agreement with Bechtel Power Corporation to obtain engineering and construction services which may be required

following an accident. Bechtel's assistance will not

be required during the early stages of the emergency

response but is more likely to be requested during

recovery activities.

A-17 REV 56 09/10

2. WESTINGHOUSE SNC has established an agreement with Westinghouse to obtain general services related to nuclear steam supply system (NSSS) operations during and following

an accident situation. Westinghouse provides a capability to respond on a 24-hour-a-day basis.

3. VOLUNTARY ASSISTANCE GROUP SNC is a signatory to two comprehensive agreements among electric utility companies: the Nuclear Power Plant Emergency Response Voluntary Assistance

Agreement and the Voluntary Assistance Agreement By

and Among Electric Utilities Involved in

Transportation of Nuclear Materials. Copies of

these agreements are included in the INPO Emergency

Resources Manual. A.9 FEDERAL GOVERNMENT SUPPORT The resources of the Federal agencies appropriate to the

emergency condition would be made available in accordance

with the National Response Plan. The emergency director is

specifically authorized to request Federal assistance on

behalf of VEGP under the provisions of the National Response

Plan. In addition to the NRC, other agencies which may

become involved are the DOE, the Federal Emergency Management

Agency, the Environmental Protection Agency, the Department

of Health and Human Services, the Department of

Transportation, and the Department of Agriculture.

A.10 CONCEPT OF OPERATIONS The emergency preparedness program for VEGP requires the coordinated response of several organizations. The emergency

organization for VEGP is described in detail in section B of

this plan. The emergency director is the key individual in

the VEGP emergency organization; one of his nondelegable

responsibilities is the decision to notify the NRC and

authorities responsible for offsite emergency measures. The

interfaces among the emergency organizations are shown on

figure A-1.

A-18 REV 23 4/96 A.10.1 Continuous Communication Capability

The emergency director will initiate the activation of the

emergency response organization by contacting the states of

Georgia and South Carolina, counties within the plume

exposure pathway EPZ, the SRS, and the NRC. All these

organizations can be contacted 24 h a day. The state and

local agencies have continuously manned communication links

for the purpose of receiving notification of a radiological

emergency. The SRS is a continuously operating facility and

can be contacted at all times. The Federal agencies which

may be requested by VEGP to provide assistance can be

notified by contacting the NRC on a dedicated communication

link, the Emergency Notification System (ENS) line.

A.10.2 State of Georgia and Burke County Operations

The State of Georgia and Burke County response is conducted

in accordance with the following framework as presented in

the Georgia Radiological Emergency Response Plan:

1. DNR is the lead State Agency for responding to peacetime radiation emergencies. The DNR radiation

emergency coordinator (REC) is responsible for

directing DNR's response. The response required may

be of varying magnitudes depending upon the nature

of the emergency.

2. In the case of occurrences of limited severity and complexity, direction and control of response and

recovery operations will be assumed by the DNR REC;

GEMA will be kept informed of conditions in order to

facilitate GEMA response and EOP activation as deemed necessary.

3. When necessitated by the magnitude and severity of an occurrence, GEMA will activate the EOP and coordinate overall response and recovery operations, with the DNR REC coordinating radiation protection

activities through the State disaster coordinator (GEMA). 4. The State is notified of an emergency at VEGP by the VEGP emergency director who has a plant communicator

call the GEMA EOC communications center on the

emergency notification network telephone. The GEMA

EOC 24-h duty officer notifies the GEMA director and

the DNR radiation emergency coordinator. Each of

these individuals can cause the declaration of an

emergency and activation of the EOC. The GEMA

director has authority to activate directly; the REC

may recommend activation to the

A-19 REV 35 6/02 Governor or the GEMA director who makes the emergency declaration. In either case, the DNR radiation emergency coordinator is responsible for

the technical direction of the offsite radiological

response. If the EOC has been activated, he reports

to the State disaster coordinator (GEMA). Recommendations and decisions to declare a state of emergency are based upon DNR's ability to

effectively protect the public with little

assistance, or, in contrast, the need for a

coordinated, multiagency effort.

5. As the designated agency to administer NRC Agreement State Programs, DNR will be the principal radiation

emergency response support agency due to the

probable requirements for special techniques, equipment, and expert personnel.

6. To the extent available, local resources, personnel authority, and emergency plans will be employed in

response to radiation emergencies.

7. When requested to assist in response and recovery efforts to radiation emergencies, personnel from

local and other State agencies will normally be

expected to perform functions and activities in

which they have expertise but may perform limited

radiation safety functions under the guidance of the

DNR REC.

The organizational structures for State and county operations

are illustrated on figures A-2 and A-3, respectively. The

Georgia Radiological Emergency Response Plan and Annex D to

the Plan provide the bases for 24-h-per-day radiological

emergency response capability for extended periods.

A.10.3 State of South Carolina and County Operations

The state of South Carolina and county response is conducted

in accordance with the following framework as presented in

the South Carolina Operational Radiological Emergency

Response Plan:

1. The Emergency Management Division (EMD) is responsible for establishing and directing the State

EOC; coordinating offsite support from state, federal, and other support agencies; determining and

directing protective actions, and providing for a

24-h notification capability.

A-20 REV 38 05/05

2. The Department of Health and Environmental Control (DHEC) is responsible for providing 24-h accident response capability, providing radiological

technical support, conducting or coordinating

offsite monitoring, and making recommendations to

EMD for protective action.

3. The state and counties are notified of an emergency at VEGP by the VEGP emergency director who has a

plant communicator call the South Carolina warning

point and the county warning points (sheriff or

central dispatchers) on the emergency notification

network. The South Carolina warning point notifies

EMD and DHEC 24-h duty officers by telephone or

beeper. The EMD then contacts the counties to ensure

warning messages have been received and contacts the

governor's office for direction, control, and

guidance. As necessary EMD alerts state response

forces, activates the state EOC and dispatches

representatives to the SNC EOF and ENC. DHEC verifies with VEGP the information received from the

South Carolina warning point; confirms the

information with EMD; and after an assessment of the

emergency situation, recommends protective measures

for the state of South Carolina and affected

county/counties, through the EMD. The Governor or

his designated representative (the Governor's

executive assistant, the director of the Governor's

Emergency Management Office, or the emergency

management coordinator) can declare an emergency and

order protective actions.

4. The SEOC will coordinate the offsite emergency response activities of State agencies, local

governments, Federal agencies, and the State of

Georgia. The director of EMD will be in charge of

SEOC. 5. If the immediacy of the emergency is such that the SEOC is not yet operational, and there is

insufficient time for DHEC to confirm the VEGP's

assessment, local government will initiate those

protective actions recommended by VEGP.

6. To the extent available, local resources will be employed in response to emergencies at VEGP. The

State arranges for emergency response capabilities

that are not available at the local government

level. 7. The State and county governments will activate their radiological emergency response plans as warranted

by the emergency action levels.

A-21 REV 24 3/97 The organizational structure for State operations is illustrated on figure A-4. The emergency response

organizations of each of the counties of Aiken, Barnwell, and

Allendale are presented in figures A-5, A-6, and A-7, respectively.

A.10.4 Savannah River Site

The DOE-SR will provide the necessary response within the SRS

reservation in accordance with the SRS Emergency Plan. The

DOE will exercise overall responsibility, jurisdiction, and

authority for conducting on-plant response operations to

protect the health and safety of SRS personnel. DOE will

provide for emergency notification and, as needed, evacuation, monitoring, decontamination, and immediate life

saving medical treatment of non SRS personnel on plant. DOE

will also provide access control for SRS areas.

DOE will provide initial radiological monitoring and

assessment support to the State of South Carolina under the

DOE Radiological Assistance Program (RAP). This includes

projected release dispersion information and offsite

radiological monitoring and assessment assistance. SRS will

also coordinate public affairs activities with the State of South Carolina, SNC and GPC.

By memorandum of agreement between DOE-SR and GPC, as assigned to SNC (see appendix 5), DOE will provide radiological monitoring within about 10 miles of VEGP in the

State of South Carolina.

REV 35 6/02 TABLE A-1 (SHEET 1 OF 2)

RESPONSIBLE INDIVIDUALS OF PRIMARY RESPONSE ORGANIZATIONS Organization Individual in Charge of Emergency Response Vogtle Electric Generating Plant Emergency director State of Georgia Governor Georgia Emergency Management

Agency State disaster coordinator Georgia Department of Natural

Resources Radiation emergency

coordinator Burke County, Georgia Chairman, Burke County Board

of Commissioners Burke County Emergency

Management Agency Emergency Management Agency

director State of South Carolina Governor South Carolina Department of

Health and Environmental

Control Manager, Nuclear Emergency

Planning Section South Carolina Office of the

Adjutant General, Emergency

Management Division Emergency Management Division director Aiken County, South Carolina Chairman, Aiken County Council Aiken County Emergency

Preparedness Division Emergency Preparedness

Division coordinator Aiken County Law Enforcement

Communications Center Sheriff Barnwell County, South

Carolina Chairman, Barnwell County

Council Barnwell County Emergency

Management Agency Coordinator, Disaster

Preparedness Agency Barnwell County Sheriff's

Office Sheriff Allendale County, South

Carolina Chairman, Allendale County

Council REV 14 12/91 TABLE A-1 (SHEET 2 OF 2)

Organization Individual in Charge of Emergency Response Allendale County Emergency

Preparedness Agency Director, Emergency Preparedness Agency Allendale County Sheriff's

Office Sheriff Department of Energy, Savannah River Manager, DOE-SR

FORMAL INTERFACES AMONGEMERGENCY ORGANIZATIONSVogtle Electric Generating PlantFIGURE A-1EMERGENCYDIRECTORSTATE OF SOUTH CAROLINAVEGPSTATE OF GEORGIAFEDERAL AGENCYSC-EMD/DHECDOE-SROTHERSTATEAGENCIESAIKEN/BARNWELL/ALLENDALECOUNTIES EPALOCALORGANIZATIONSEOFDIRECTION OR COORDINATION PATHINFORMATION PATHAID OR ASSISTANCE PATHGEMA NRCDNR/EPDOTHERSTATEAGENCIESBURKE COUNTY EMALOCALORGANIZATIONS(IN ACCORDANCEWITH THE FRERP)REV 35 6/02 GOVERNORSTATEDISASTERCOORDINATORDEPARTMENT OFNATURAL RESOURCESENVIRONMENTAL PROTECTION DIVISIONDEPARTMENT OFPUBLIC SAFETYGEORGIAEMERGENCYMANAGEMENTAGENCY (GEMA)DEPARTMENT OFTRANSPORTATIONDEPARTMENT OFHUMANRESOURCESDEPARTMENT OFAGRICULTUREFORESTRYCOMMISSIONDEPARTMENT OFADMINISTRATIVESERVICESOTHERAGENCYRESOURCESPolice servicesPublic informationCommunication and warningSearch and rescue evacuation State military support Disaster assistance centerDirection and control damage assessmentPublic property assistanceCoordination of trainingCoordination of volunteersEngineering servicesTransportation servicesCommunicationHealth and social servicesContaminated food and milk controlCoordination with USDAFire controlTransportation servicesHeavy equipmentCommunicationsProcurementTransportation servicesCommunication equipmentVogtle Electric Generating PlantFIGURE A-2STATE OF GEORGIA OPERATING IN A DECLARED RADIOLOGICAL EMERGENCYREV 22 5/95 Coordination Link Operational Link VEGPLEGEND:OPERATIONAL LINKCOORDINATION LINKBURKECOUNTY EMA GEMAFIREDEPARTMENTSHERIFF'SDEPARTMENTBOARD OFEDUCATIONMUNICIPALPOLICEVOLUNTEERORGANIZATIONSCOUNTYPUBLICWORKSMUNICIPALPUBLIC WORKSCOUNTY HEALTHDEPARTMENTBURKE COUNTYHOSPITALDEFACSCHAIRMANBURKE COUNTYCOMMISSIONVogtle Electric Generating PlantFIGURE A-3OPERATIONAL RELATIONSHIPS AMONG BURKE COUNTY RESPONSE ORGANIZATIONSREV 2 11/85 GOVERNORDirection, Control andGuidanceResource ManagementState Military SupportPlanningResponseDamage AssessmentPreparednessOperationsEMERGENCY MANAGEMENT DIVISIONPublicInformationDHECNUCLEAR EMERGENCYPLANNING SECTIONRadiological Assessment and EvaluationProtective MeasuresDamage AssessmentHealth ServiceSTATE LAWENFORCEMENTDIVISIONDEPARTMENT OF HIGHWAYS AND PUBLIC TRANSPORTATIONFORESTRYCOMMISSIONDEPARTMENT OFSOCIALSERVICESWILDLIFE ANDMARINERESOURCESDEPARTMENTCLEMSONUNIVERSITYEXTENTIONSERVICEDEPARTMENT OFAGRICULTUREOTHERAGENCYRESOURCESAlerting and WarningPolice ServiceCommunicationSearch and RescueDebris RemovalAlerting and WarningCommunication Traffic ControlTransport ServiceDamage AssessmentSecurityFire Control ServiceHeavy EquipmentSearch and RescueCommunicationTransport ServiceEmergency Welfare ServiceSearch and RescueCommunicationTransport ServiceAgricultural Data and InformationContaminated Food and Milk ControlFood and Crop Interdiction Coordination with USDA Vogtle Electric Generating PlantFIGURE A-4STATE OF SOUTH CAROLINA OPERATINGIN A DECLARED RADIOLOGICAL EMERGENCYREV 35 6/02 OTAG EMERGENCY MANAGEMENT DIVISION CHAIRMANAIKEN COUNTYCOUNCILLEGEND:OPERATIONAL LINKCOORDINATION LINKAIKENCOUNTYADMINISTRATORVEGPSOUTH CAROLINAEMERGENCYMANAGEMENTDIVISIONAIKEN COUNTYEMERGENCYPREPAREDNESSDIVISIONSHERIFF'SOFFICE FIREDEPARTMENTEMERGENCYSERVICEDEPARTMENT OFSOCIALSERVICESLAWENFORCEMENTCOMMUNICATIONCENTERPUBLIC WORKSDEPARTMENTEMERGENCYMEDICALSERVICESSCHOOLDISTRICTMUNICIPALITIESRESCUE/AMBULANCEVOLUNTEERORGANIZATIONS REDCROSSVogtle Electric Generating PlantFIGURE A-5AIKEN COUNTY, S.C.EMERGENCY RESPONSE ORGANIZATIONREV 35 6/02

LEGEND:OPERATIONAL LINKCOORDINATIONL LINKCHAIRMANBARNWELLCOUNTY COUNCILCHAIRMANBARNWELLADMINISTRATORAGENCYSOUTH CAROLINAEMERGENCYMANAGEMENTAGENCYVEGPSHERIFF'SOFFICEFIREDEPARTMENTEMERGENCYSERVICEDEPARTMENT OFSOCIALSERVICESVOLUNTEERORGANIZATIONCOMMUNICATIONSERVICEEMERGENCYMEDICALSERVICESSCHOOLDISTRICTMUNICIPALITIESRESCUEAMBULANCESERVICEFIGURE A-6Vogtle Electric Generating PlantBARNWELL COUNTY, S.C.EMERGENCY RESPONSE ORGANIZATION R E V 3 5 6/0 2BARNWELL COUNTYMANAGEMENTEMERGENCY CHAIRMANALLENDALECOUNTY COUNCILLEGEND:OPERATIONAL LINKCOORDINATION LINKALLENDALECOUNTYADMINISTRATORVEGPSOUTH CAROLINAEMERGENCYMANAGEMENTDIVISIONAIKEN COUNTYEMERGENCYPREPAREDNESSDIVISIONSHERIFF'S OFFICE FIREDEPARTMENTEMERGENCYSERVICEDEPARTMENT OFSOCIALSERVICES LAWENFORCEMENTCOMMUNICATIONCENTERDEPARTMENTOF ROADS ANDBRIDGESEMERGENCYMEDICALSERVICESOFFICE OFSCHOOLSUPERINTENDENTMUNICIPALITIESRESCUE/AMBULANCEVOLUNTEERORGANIZATIONSREDCROSSVogtle Electric Generating PlantFIGURE A-7ALLENDALE COUNTY, S.C.EMERGENCY RESPONSE ORGANIZATIONREV 35 6/02

B-1 REV 61 06/14 B. VEGP EMERGENCY ORGANIZATION Initial staffing of the onsite emergency organization will be provided from personnel normally employed at the site. For

reference throughout this section, the organizational chart

for the Vogtle Electric Generating Plant (VEGP) staff is

presented in figure B-1. If the need arises, this staff will

be augmented substantially by the addition of Southern

Nuclear Operating Company (SNC) personnel and by personnel

from other organizations. This section includes a

description of the emergency duties of the normal shift

complement, a discussion of the manner in which emergency

assignments are to be made, a listing of additional support

personnel on whom VEGP can rely, and a description of the

relationships between onsite and offsite response activities.

B.1 NORMAL PLANT ORGANIZATION The organizational structure shown on figure B-1 represents the pool of personnel available on site during normal working

hours. Approximately 700 people are stationed at the site

during the standard workday.

The normal operating crew for two units includes a shift

supervisor, licensed plant operators, and non-licensed plant

operators. A shift manager is also on shift during operation (as defined in the Technical Specifications). Personnel from

the Chemistry and Health Physics, Maintenance, and Security

Departments are also on site continuously. Refer to table B-

1 of this section for minimum staffing requirements.

B.2 EMERGENCY ORGANIZATION The emergency director has the responsibility to classify an event in accordance with the emergency classification system (described in section D). Classification of an event into one

of the four emergency categories (Notification of Unusual Event (NUE), Alert, Site Area Emergency, or General Emergency)

activates the VEGP emergency organization. The extent to which

the emergency organization is activated depends on the severity

of the situation. Table B-1 provides a summary of personnel

available on shift and those who would be available within 60

min of notification. A copy of the On-Shift Staffing Analysis which forms the technical basis for Table B-1 Minimum Shift Staffing is maintained in the SNC document management system.

Reference OSA-VEGP-001.

For an NUE, the emergency director assigns responsibility for

making the appropriate notifications and directing the proper

response; but no further activation of the emergency

organization is required.

If the event is classified as an Alert, the technical support

center (TSC), operations support center (OSC), and Emergency

Operations Facility (EOF) will be activated.

B-2 REV 61 06/14 For this classification, the emergency organization is structured as shown on figure B-2. The corporate resources

and operation is presented in appendix 7, Emergency

Operations Facility. Corporate personnel who report to the

plant site will be integrated into the VEGP emergency

organization.

For a Site Area Emergency or General Emergency, the emergency

organization and EOF will be fully activated. The

organization will be as shown in figure B-3.

During hostile action, ERO members would likely not have

access to the onsite emergency response facilities. A

security related emergency may delay the ordering of facility

activation in order to protect plant personnel from the

security threat. The decision to delay activation of the

facilities will be made by the Emergency Director. In such cases offsite ERO personnel will be directed to an alternative facility to minimize delays in overall site response by permitting ERO assembly without exposing responders to the danger of hostile action.

Relationships among the VEGP emergency organization and other

elements of emergency response are shown on figure A-1.

B.2.1 Emergency Organization Responsibilities

Following an Alert or higher emergency declaration, the positions shown on figures B-2 and B-3 will be

filled by VEGP or SNC personnel as discussed below.

1. EMERGENCY DIRECTOR Plant personnel that may be designated as emergency directors are listed in table B-2 of this plan.

They will receive training as specified in table 0-2

of this plan prior to becoming qualified to fill

this position. Their nonemergency positions provide

them plant knowledge and supervisory skills

necessary to fill the emergency director position. The emergency director has the authority, management ability, and knowledge to assume the overall

responsibility for directing VEGP staff in an

emergency situation. Initially this position will

be filled by the shift manager or the shift

supervisor if the shift manager can not be located

expeditiously. The responsibility for emergency

direction will be transferred to qualified senior

management personnel designated in procedure 91101-C

after receiving an appropriate B-3 REV 57 06/12 briefing and becoming familiar with the current status of events. Turnover with the accompanying briefing will include, but is not limited to, the following:

1. Review of logs and status boards.

2. Discussion with the incumbent including emergency classification, summary of events, offsite notifications, plant status, equipment status, outstanding orders, any noted deficiencies and

completed checklist items.

3. Discussion with staff, as needed. Following relief, announcement will be made to staff of the transfer of responsibility. (See Procedure 91102-C for specifics.) The primary

and alternates for the position of emergency

director are shown on table B-2. The emergency director manages the following activities for the duration of the emergency:

• Notification and communication:

directs the notification of VEGP, SNC and GPC personnel

and notifies and maintains open communications

with offsite authorities regarding all aspects

of emergency response.

• Emergency response facilities:

oversees the activation and staffing and requests

additional assistance, as needed.

• Emergency operations:

has authority over those actions taken to mitigate the emergency

condition or reduce the threat to the safety

of plant personnel or the public, including

the recommendation of protective actions to

offsite authorities.

• Emergency services:

provides overall direction for management of procurement of

site needed materials, equipment, and

supplies; documentation; accountability; and

security functions.

• Emergency operations planning:

provides overall direction for the management of

planning for procedure, equipment, and system

development to support emergency operations.

• Discretionary authority:

can modify emergency implementing procedures; may tailor the

emergency organization to fit the specific

staffing needs on a case by case basis. The emergency director may not delegate the following responsibilities:

B-4 REV 57 06/12

• The decision to notify offsite emergency response agencies.

• The decision to recommend protective actions to

offsite authorities.

• Declaration of emergency classifications.

• Authorization for plant personnel to exceed 10

CFR 20 radiation exposure limits.

• The decision to terminate the emergency.

• Request for Federal assistance.

• The decision to order site dismissal of non-

involved personnel from the site at an Alert

classification level.

• The decision to order non-involved personnel to

proceed to a reception center and receive

radiological monitoring. The emergency director may operate from the control room or TSC at his discretion. He may

act as the TSC manager during the early phases of

emergency response until the EOF is activated. It

is the intent of SNC that the ED will be

transferred from the Control Room as soon as

practicable.

2. TSC STAFF

a. TSC Manager The TSC manager performs the following activities:

• Coordination of inputs and recommendations

from technical and corrective action

advisors.

• Direction of onsite emergency personnel

involved in restoration of the plant to a

safe condition.

• Technical assistance and operations guidance

to control room personnel.

• Direction of TSC staff in analysis of

problems, design and planning for temporary

modifications, and development of temporary

emergency operating procedures.

• Recommendation of protective actions to the

emergency director based on plant

conditions.

• Providing recommendations on emergency

classifications to the emergency director.

B-5 REV 37 4/03

b. TSC Support Coordinator The TSC support coordinator directs the clerical and logistic activities in the TSC.

He ensures that support staff, including

clerks, status board keepers, and

communicators, are available in sufficient

numbers and that office supplies, drawings, and

other documents are available to TSC and OSC

personnel. He is responsible for timely

completion of offsite notification. He ensures

that transportation and communication needs are

satisfied. He arranges for additional offsite

support personnel and equipment working in

conjunction with the EOF support coordinator.

(Designees are identified on table B-2).

c. Engineering Supervisor The engineering supervisor directs a staff of engineers with expertise in reactor

engineering, thermal and hydraulic analysis, instrumentation and control, and mechanical and

electrical systems. He directs the analysis of

plant problems, core damage, and provides

recommendations for plant modifications to

mitigate the effects of the accident.

d. Maintenance Supervisor The maintenance supervisor manages the planning and coordination of repair, damage control, and

plant modification activities. He works

closely with the engineering supervisor in

planning for plant modifications and repairs.

e. Operations Supervisor The operations supervisor analyzes problems associated with systems operations and provides

recommendations for procedures for mitigating

the emergency situation.

f. Health Physics Supervisor The health physics supervisor is responsible

for onsite and in-plant radiological controls.

He provides guidance to the maintenance

supervisor related to radiological

considerations associated with plant

modification and repair and provides direction

to the OSC manager related to the health

physics controls for emergency teams. He

performs offsite dose assessment prior to EOF

activation and keeps the dose assessment

supervisor in the EOF informed of the radiological status of the plant.

B-6 REV 37 4/03

g. Chemistry Supervisor The chemistry supervisor is responsible for directing and evaluating in-plant chemistry and analyses, directing and evaluating post

accident sampling, and assisting in core damage

assessment.

h. TSC Security Supervisor The TSC security supervisor coordinates the security functions including accountability and

site access control. He coordinates the

processing of offsite personnel who require

authorization to enter the site. When directed

by the emergency director, will request

assistance from civil law enforcement

authorities, as required.

3. OSC STAFF

a. OSC Manager The OSC manager receives direction from the TSC personnel to dispatch emergency teams (e.g.,

firefighting, search and rescue, first aid, repair, etc.) to prescribed areas of the plant

or site. The OSC manager directs composition

of the teams to ensure that appropriately

qualified personnel are assigned. In

particular, he will ensure that proper health

physics coverage is provided. The OSC manager

will provide specific instructions to the team

leaders. He will also maintain communications

with the teams that remain assigned to the OSC

and monitor the status of their activities.

b. OSC Personnel Selected emergency response personnel will report to the OSC as directed. Depending on the

nature of the emergency, personnel from the

Maintenance, Operations, Chemistry and Health

Physics Departments will be directed to report

to the OSC. The following emergency teams will

be formed as necessary:

• Backup fire brigade.

• Search and rescue.

• First aid.

• Damage assessment.

• Damage control.

B-7 REV 61 06/14

• Repair and modification.

• Field monitoring. Each team will be headed by a designated team leader, who will maintain communications with the OSC, TSC, or EOF.

4. EOF STAFF

• The EOF staff is described in Appendix 7

5. ALTERNATIVE FACILITY STAFF The ERO staff will be directed to report to the Alternative Facility during a security related event, or other events that preclude onsite access. This facility functions as a staging area for augmentation of emergency response staff and provides the capability for communication with the EOF, Control Room, and plant security.

From this facility the ERO will support event response by performing engineering assessment activities, including damage control team planning and preparation for return to the site, the command and control function will remain with the ED in the control room until relieved by another onsite ED. Dose assessment and offsite notifications will be performed by the EOF.

B-8 REV 38 05/05

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B-9 REV 56 09/10 B.2.2 Emergency Organization Assignments

Table B-2 identifies by title the individuals who will fill

the key emergency positions.

A sufficient number of people are identified to ensure that

all emergency positions on table B-2 will be filled.

B.2.3 Other Support Services

1. CONTRACTOR SUPPORT Arrangements have been made to obtain support services from Bechtel Power Corporation and Westinghouse, if required. These organizations will

initially be contacted by the EOF support

coordinator to arrange for the required assistance.

2. MEDICAL ASSISTANCE Agreements are in place with Tetra Tech, Burke Medical Center, Doctors Hospital, and Burke County Emergency Management Agency to provide assistance and training for injured personnel, including cases involving radioactive contamination. This assistance will be requested whenever necessary in

accordance with plant procedures.

B-10 REV 19 1/94

3. AGENCY SUPPORT Assistance may be requested from Burke County, the State of Georgia, or Federal agencies. Section A of this Plan describes the assistance that may be

requested. Any requests for aid will be made by the

emergency director. B.2.4 Interfaces Among Response Groups

Section A, figure A-1, illustrates the integrated

organization for response to an emergency at VEGP.

REV 60 07/13 TABLE B-1 (SHEET 1 OF 2)

MINIMUM STAFFING FOR POWER OPERATION Major Functional Area Major Tasks Position Title or Expertise On Shift Augmen-tation in 60 min Plant operations and assessment of operational aspects Shift manager (SRO)

Shift supervisor (SRO)

Plant operator (RO)

System operator Shift Support Supervisor (SSS) 1(a) 2(a) 4(a) 7 (a)(e)(f) 1 -

-

-

- Emergency direction and control (emergency director)

Overall management of

emergency organization Shift supervisor;

shift manager 1(b) - Notification/communication Notification of VEGP, State, local, and Federal personnel Shift trained personnel 2(b) 2 Radiological accident assessment

and support of operational

accident assessment EOF direction

Offsite dose assessment Corporate Management;

HP/Chemistry Shared foreman -

1 1

- Offsite surveys onsite (out of plant)

Chemistry technicians, HP technicians, System operator or other trained personnel 2 3 FMT Communicator Maintenance Supervisor 1 - Chemistry/radiochemistry Chemistry technicians or equivalent 1 1 Plant system engineering, repair

and corrective actions Technical support (including

core/thermal hydraulics)

Shift technical advisor or

engineer 1(c) - Electrical Mechanical -

- 1 1 Repair and corrective actions Mechanical maintenance System operator

Electrical maintenance

Instrumentation and control

technician 1

-

1 1 1 1

1

- Protective actions (in plant) Radiation protection:

• Access control

• HP coverage for repair, corrective actions, search

and rescue, first aid and

fire fighting

• Personnel monitoring

• Dosimetry

• Decontamination

• In-plant Survey Health Physics technicians or

other trained personnel 2 2

REV 60 07/13 TABLE B-1 (SHEET 2 OF 2)

MINIMUM STAFFING FOR POWER OPERATION Major Functional Area Major Tasks Position Title or Expertise On Shift Augmentation in 60 min Firefighting Fire brigade per FSAR Local support Rescue operations and first aid 2(b) Local support Site access control and

personnel accountability Security, firefighting

communications, personnel

accountability Security personnel Per Security Plan Totals 24(d) 14

a. Refer to technical specifications for nonpower operation. b. May be provided by shift personnel assigned other functions.

c. Required unless shift manager or the individual with a senior operator license meets the qualification for the STA as required by the NRC. d. Does not include positions footnoted with a (b) or (c).

e. Four of the seven system operators may be assigned to the Fire Brigade.

f. One SO may be assigned to FMT.

REV 60 07/13 TABLE B-2 (SHEET 1 OF 2)

EMERGENCY ORGANIZATION ASSIGNMENTS Emergency Position Designees Emergency Director Operations Director; Maintenance Manager; Work Management Director; Engineering Director; Manager of Site Projects; Engineering Programs Manager or other qualified senior management personnel.

EOF Staff Corporate Staff as defined in Appendix 7 TSC Manager Supervision from onsite staff as designated in procedure 91101-C TSC Support Coordinator Onsite staff as designated in procedure 91101-C TSC Engineering Supervisor Supervision from onsite staff as designated in procedure 91101-C TSC Maintenance Supervisor Supervision from onsite staff as designated in procedure 91101-C TSC Operations Supervisor Supervision from onsite staff as designated in procedure 91101-C TSC Health Physics Supervisor Supervision from onsite staff as designated in procedure 91101-C TSC Chemistry Supervisor Supervision from onsite staff as designated in procedure 91101-C Engineers Plant engineers TSC Security Supervisor Supervision from onsite staff as designated in procedure 91101-C REV 43 05/06 TABLE B-2 (SHEET 2 OF 2)

Emergency Position Designees OSC manager Supervision from on-site staff as designated in procedure 91101-C Dosimetry team Qualified health physics personnel Communicators (Offsite for

Control Room, TSC and EOF)

Plant engineers; nuclear specialist; administrative assistants; operations

personnel, corporate staff Clerks Administrative assistants In-plant radiation monitoring

team Selected emergency response personnel Post accident sampling team Selected emergency response personnel Damage control/assessment/

repair team Selected emergency response personnel Search and rescue team Selected emergency response personnel Backup fire brigade Selected emergency response personnel First aid team Selected emergency response personnel Field monitoring team Selected emergency response personnel Shift supervisor

Plant operators

Plant system operators Normal operating shift personnel

EMERGENCYDIRECTOREOF STAFFAS DESCRIBED INAPPENDIX 7 OF THIS PLAN TSCMANAGERSHIFTMANAGER(CONTROL ROOM)SHIFTSUPERVISOROSCSYSTEMOPERATORSPLANTOPERATORSTSC SUPPORT COORDINATORTSC SECURITY SUPERVISORTSC HEALTH PHYSICS SUPERVISORTSC CHEMISTRY SUPERVISOR TSC MAINTENANCE SUPERVISOR TSC OPERATIONS SUPERVISOR TSC ENGINEERING SUPERVISORVogtle Electric Generating Plant FIGURE B-2RESPONSE ORGANIZATION FOR ALERTREV 43 05/06TSC ENGINEERS

FEDERAL, STATEAND OTHER OFFSITEAGENCIESCOORDINATIONAUTHORITYEMERGENCYDIRECTORTSCMANAGEREOFMANAGEROSCMANAGERSHIFTMANAGERSHIFTSUPERVISORPLANTOPERATORSDAMAGE CONTROL/ASSESSMENT/REPAIR TEAMSEARCH AND RESCUE TEAMBACKUP FIRE BRIGADEFIRST AIDHEALTH PHYSICS SUPERVISOR--IN-PLANT MONITORING TEAMOPERATIONS SUPERVISORCHEMISTRY SUPERVISORMAINTENANCE SUPERVISORENGINEERING SUPERVISORTSC SECURITY SUPERVISORTSC SUPPORT COORDINATORREV 43 05/06SITE AREA OR GENERALEMEGENCY RESPONSE ORGANIZATION FIGURE B-3SYSTEMOPERATORSVogtle Electric Generating PlantEOF STAFF AS DESCRIBED IN APPENDIX 7 OF THIS PLAN

C-1 REV 61 06/14 C. EMERGENCY RESPONSE SUPPORT AND RESOURCES C.1 STATE AND LOCAL GOVERNMENTAL SUPPORT The State of Georgia through the Georgia Emergency Management Agency (GEMA) has the lead agency responsibility for

responding to emergency situations throughout Georgia. Under

the procedure established by the Georgia Emergency Operations

Plan (EOP), which was developed pursuant to the Governor's

Executive Order, the Department of Natural Resources (DNR)

radiological emergency response team in coordination with

GEMA assesses the radiological conditions at the site of an

incident and determines whether a state of emergency should

be declared. The Governor of the State of Georgia or GEMA

can declare an emergency based upon the assessment made by

GEMA and/or DNR.

The South Carolina agencies responsible for responding to a

radiological emergency are the Office of the Adjutant

General, Emergency Management Division (EMD) and the

Department of Health and Environmental Control (DHEC). The

Nuclear Emergency Planning Section within DHEC is given the

primary role in responding to the technical aspects of a

nuclear accident: assessment of the radiological consequences

and provision of protective action guidelines to state and

local authorities, through the EMD. The EMD is assigned

primary responsibility for the planning effort and for

coordinating the state response operations. Jointly with

DHEC, EMD determines whether a state of emergency should be

declared. Upon EMD's advising the Governor of South Carolina

that a radiological emergency exists, the Governor declares

an emergency.

Agreements are in place with the State of Georgia, Burke County, Georgia, and Aiken, Allendale, and Barnwell Counties in South Carolina, to provide available resources and equipment to support the mitigation and response to an emergency at Plant Vogtle to include Hostile Action Based events. These resources include, but are not limited to, Local Law Enforcement Agency (LLEA) assets, Firefighting assets, medical support resources (including transportation) and coordination through an Incident command Post. Requests for offsite resources and equipment will be communicated from the control room to the Burke County 911 Center or through the Incident Command Post as applicable based on the nature and timing of the event. Copies of these agreements are maintained in the SNC document management system and are included by reference in Appendix 2.

The concept of the operations for which the State and local

governments are responsible, together with a discussion of

responsibilities assigned to various state/county

governmental agencies are contained in the respective

Radiological Emergency Response Plan. For a complete

discussion of authority, assigned responsibilities, C-2 REV 61 06/14 capabilities, and activation and communication arrangements, refer to these plans.

Representatives from the states of Georgia and South Carolina

will be dispatched to the emergency operations facility (EOF)

and the Joint Information Center (JIC). As requested, Vogtle

Electric Generating Plant (VEGP) will send representatives to

the offsite government centers. Table C-2 lists the state

and county EOCs with their addresses.

C.2 FEDERAL GOVERNMENTAL SUPPORT In addition to coordination with State/county governmental entities in an emergency situation, VEGP may require

assistance from certain Federal agencies in the areas of

communications, radiological monitoring and laboratory

analysis, transportation, and disaster relief. Requests for

Federal assistance are directed as needed by the emergency

director, and usually these requests are channeled through

GEMA. The exceptions to this procedure are direct contacts

between the VEGP Emergency Organization, the Nuclear

Regulatory Commission (NRC), and Department of Energy, Savannah River (DOE-SR).

In the event of an incident in which Federal assistance is

needed to supplement county and State emergency response

capabilities, principal points of contact for State

government are as follows:

• The Federal Emergency Management Agency (FEMA),

Region Headquarters in Atlanta.

• The Department of Energy (DOE), Region Operations

Office in Aiken, South Carolina.

• The Environmental Protection Agency (EPA), Region

Headquarters in Atlanta.

FEMA is assigned lead responsibility for Federal offsite

nuclear emergency planning and response (per Title 44 CFR

351). FEMA is also delegated responsibility for development

and promulgation of the Federal Radiological Emergency

Response Plan (FRERP). The FRERP assumes that states will be

responsible for overall management of offsite emergency

response. The Federal government's role consists of

providing technical and/or logistical resource support at the

request of State emergency management.

Federal emergency response consists of technical and

nontechnical components. The NRC and FEMA jointly coordinate

federal emergency response actions. The NRC coordinates

technical aspects, and FEMA coordinates nontechnical aspects

of Federal response.

C-3 REV 61 06/14 The NRC and FEMA have stated that they each expect to have representatives at the Vogtle Electric Generating Plant (VEGP) within about 3 h after receiving notification. DOE

can give assistance within about 2 h.

GEMA has the overall responsibility for making arrangements

to support the Federal offsite response. County emergency

management directors in whose area Federal government

response personnel are to be housed or whose county is

otherwise affected by the Federal response will cooperate

with the Federal government and GEMA in planning for and

making the necessary support arrangements.

Suitable quarters for lodging Federal government personnel

include commercially available hotel/motel accommodations, dormitory, military billets, or similar accommodations. Such

facilities exist in the greater Augusta area. Twenty-four

hour food service is also available in the area. Commercial

auto rentals are available at Bush Field and in the Augusta

area.

Within several hours of notification, Federal response

personnel will begin arriving at or near VEGP. A Federal

Response Center (FRC) will be established for the

coordination and exchange of information among various

Federal agencies during an emergency at VEGP. The Federal

Radiological Monitoring and Assessment Center (FRMAC) will be

established for the coordination of Federal monitoring and

assessment assistance with State and local governments.

Upon activation of the Federal Radiological Emergency Response Plan (per 44 CFR 351), DOE is tasked with providing

telecommunications support to Federal agencies assisting in

offsite radiological monitoring. The DOE has a

communications pod which has telephone, VHF radio, microwave, and video receiving and transmission capability. Up to 20

telephone lines may be initially installed. The pod has

radio-to-telephone patch capability and microwave capability

for about 60 miles. The DOE has written agreements with all

telephone companies to provide additional telephone

communications, including satellite capability, within 24 to

48 hs. This capability will be used to supplement

communications among the FRC, FRMAC, EOF, and States of

Georgia EOC/FEOC and South Carolina EOC.

Airfields in the plant vicinity that may be used to support the Federal response, as well as that of other response

groups, include a commercial airport with scheduled service

and nearby municipal airports that can accommodate small

aircraft.

Characteristics of four selected airports in the vicinity of VEGP are presented in Table C-1. Bush Field (Augusta, Georgia) is the closest major airport capable of providing

services for large aircraft. The field is a scheduled

commercial air carrier facility capable of handling large C-4 REV 61 06/14 multiengine jet aircraft. It is also a military air headquarters for the U.S. Army operating out of Ft. Gordon, Georgia.

Daniel Field (Augusta, Georgia) and Aiken Municipal (Aiken, South Carolina) are air fields capable of servicing and

maintaining medium-size jet and propeller aircraft. The

Burke County Airport (Waynesboro, Georgia) is the nearest

airport to the VEGP and is used only by small general

aviation aircraft.

C.3 VEGP SUPPORT VEGP provides space, telephone communications, and administrative services for NRC and FEMA personnel at the TSC

and EOF. Up to five NRC representatives can be accommodated

at the TSC. In the EOF, space is provided for nine NRC

personnel and one FEMA representative. NRC representatives

may also be present in the control room. Emergency

notification system (ENS) telephones and commercial

telephones are available in the control room, TSC, and EOF.

Health physics network (HPN) telephones are available in the

TSC and the EOF.

C.4 OTHER SUPPORT The onsite laboratory is equipped to analyze all normal in-plant samples. The equipment includes an ion chromatograph, gas chromatograph, gamma spectrometer, and other analytical

support equipment.

Field samples will be taken by VEGP field monitoring teams.

These teams will take direct radiation readings and will

collect air samples, soil samples, vegetation samples, and

water samples as directed by the dose assessment manager.

The samples will be scanned with field instrumentation and

will then be taken to VEGP for laboratory analyses. If

necessary, samples will be transported to the Georgia Power

Company (GPC) Environmental Laboratory located in Smyrna, Georgia, or to Plant Hatch for more refined analyses.

Environmental samples will be collected by corporate

personnel. These samples will be obtained from the current

fixed environmental program which is described in section I.

Environmental samples will be sent to Plant Hatch or the GPC

environmental laboratory.

In-plant samples such as effluent and air samples will be

analyzed using a gamma spectrometer located in the counting

room. Post-accident sampling is described in section I of the

plan.

Private organizations that supply engineering, health

physics, and general emergency support are listed below.

C-5 REV 61 06/14

• Southern Nuclear Operating Company (SNC), Birmingham, Alabama.

• Southern Company Services, Inc., (SCS) Birmingham, Alabama.

• Westinghouse, Pittsburgh, Pennsylvania.

• Institute of Nuclear Power Operations (INPO),

Atlanta, Georgia.

SNC serves as the primary engineering and design organization

for the plant. The nuclear steam supply systems for the plant

were purchased from Westinghouse. Westinghouse continues to

provide operations support to the company in plant

modifications, licensing, and engineering.

As a member of INPO, SNC is provided with INPO's emergency

response manual. This manual identifies the quantity of

personnel that various organizations (utilities, service

companies, and reactor vendors) could reasonably be expected

to make available in response to a request for emergency

support.

As referenced throughout this Plan, some offsite GPC and SNC

departments may be involved in the emergency response effort.

These departments have, where appropriate, developed separate

nuclear emergency response plans and procedures governing

their emergency functions. Coordination of these plans to

ensure a consistent integrated response is the responsibility

of the Corporate Emergency Planning Section. These specific

plans include:

• Corporate Emergency Plan, controlled by the SNC

Corporate Emergency Planning Section.

• Emergency Communication Plan, controlled by the GPC

Corporate Communications Department.

• VEGP Security Plan, controlled by the manager nuclear

security.

• VEGP Fire Protection Plan, controlled by engineering

support.

REV 2 11/85 TABLE C-1 AIRPORT CHARACTERISTICS (SHEET 1 OF 2)

Bush Field Augusta, Georgia Daniel Field Augusta, Georgia Aiken Municipal Aiken, South Carolina Burke Co. Airport Waynesboro, Georgia Runway length and type (Runway 17-35)

8,001 ft x 150 ft;

asphalt (Runway 05-23)

3,877 ft x 150 ft;

asphalt (Runway 06-24)

5,000 ft x 100 ft;

asphalt (Runway 07-25)

3,200 ft x 75 ft;

asphalt (Runway 08-26)

6,001 ft x 150 ft;

asphalt (Runway 10-28)

3,773 ft x 150 ft;

asphalt (Runway 18-36)

3,800 ft x 75 ft;

asphalt Runway bearing (a) Capacity (1000's lbs.) (Runway 17-35)

130(S) 166(D)

358(DT) Unknown (Runway 06-24) 22(s) 40(D) (Runway 07-25)

29(S) Runway 08-26) 52(S)

71(D) 126(DT)

(Runway 18-36) 22(s)

Approach systems (b) ILS, VOR, NDB, (R) VOR, RNA NDB VFR Fuel Aviation gas, 100 LL

Jet A Aviation gas, 100 LL

Jet A1 Aviation gas, 100 LL

Jet A Aviation gas, 100

LL Hours Attended continuously Attended daylight

hours Attended continuously Attended Monday-Saturday daylight

hours; attended

Sunday noon to dusk Runway lighting

intensity High Medium Medium Low Maintenance and

facilities Major airframe and

major power plant

repairs; hangars;

electrical hookups;

tiedowns. One hangar

is leased by the U.S.

Army and has

electrical hookups.

One or more large

travel-trailers (e.g., "Air Stream")

can be accommodated

by this hangar.

Major airframe and

major power plant

repairs; two hangars;

electrical hook-ups;

tiedowns. Each hangar

can accommodate one or

more large travel

trailers (e.g., "Air

Stream").

Major airframe and

major power plant

repairs; five

hangars(small

aircraft); electrical

hook-ups; tiedowns; 1

telephone in office, 1 in a hangar (both

extensions of the

same line).

No repairs

No hangars

No space left for

tiedowns; electrical

hook-ups; one

telephone in office REV 28 3/99 TABLE C-1 AIRPORT CHARACTERISTICS (SHEET 2 OF 2)

Bush Field Augusta, Georgia Daniel Field Augusta, Georgia Aiken Municipal Aiken, South Carolina Burke Co. Airport Waynesboro, Georgia Communications In the U.S. Army hangar; four phone lines (1 commercial, 1 autovon, 2

dedicated lines to Ft.

Gordon); two radios (VHF-

all frequencies, FM with

tactical military

frequencies).

Three telephones(1

in airport office, 1

in each of 2

hangars-all

extensions of the

same line).

One telephone in

office, 1 in a

hangar, (both

extensions of the

same line).

One telephone in

office Accommodations One motel at airport; two restaurants, 24-h day in

the area; rental cars

available Motel 3 mi. from

airport; one

restaurant; taxi One restaurant 1/4

MI from airport;

rental cars (must

radio ahead)

None at airport;

no taxi; managers

vehicle may be

available; no

rental cars;

several motels in Waynesboro; no

restaurant at

airport Remarks Commercial scheduled service; Military Air

Headquarters U.S. Army, Ft. Gordon Possible runway extension of 700 ft;

date unknown Possible runway

extension of 1000

to 1,300 ft;

possible construction at

aircraft shelters (date unknown)

Distance from

Plant Vogtle (MI) Air: 17 23 35 16 Road: 26 35 50 18 a. (S) Single-wheel type gear (DC-3). b. ILS Instrument landing system. (D) Dual-wheel type gear (DC-6). VOR VHF navigation facility, omnidirectional. (DT) Dual-tandem type gear (707). NDB Nondirectional beacon.

RNA Random area navigation.

VFR Visual flight rules.

(R) Radar. Source: Airport Facility Directory, U.S. Department of Commerce, September 26, 1985 and facility manager

REV 56 09/10 TABLE C-2 STATE AND COUNTY EMERGENCY OPERATION CENTERS (EOC)

Organization EOC Location Georgia State EOC(SEOC) 935 E. Confederate Ave.

Atlanta, GA Georgia Forward EOC Burke County EMA Building 277 GA Hwy 24 and Perimeter Road Waynesboro, GA Burke County EOC Burke County EMA Building

277 GA Hwy 24 and Perimeter Road Waynesboro, GA South Carolina SEOC 1100 Fish Hatchery Road West Columbia, SC 29172 Aiken County EOC 420 Hampton Ave, NE Aiken County Sheriff's Department Building Aiken, SC 29801 Allendale County EOC 292 Barnwell HWY Allendale, SC 29810 Barnwell County EOC Barnwell County Office 57 Wall Street Barnwell, SC

D-1 REV 48 03/08 D. EMERGENCY CLASSIFICATION SYSTEM D.1

SUMMARY

OF EMERGENCY CLASSIFICATION LOGIC The classification system for the Emergency Plan for the Vogtle Electric Generating Plant (VEGP) is based on four

emergency classes. The classes determine initial steps

to be taken by VEGP on site and by corporate emergency

response elements and the actions which are taken by the

offsite support organizations. An emergency

classification is indicative of the status of the plant.

Inputs to the emergency classification system include

the status of various plant systems, radiation levels in

and around plant areas, and the rate of release of

radioactivity from the plant.

The emergency classes are used by offsite authorities to

determine which of the preplanned actions to be taken by

their emergency organizations. Protective actions taken

on behalf of members of the public are the

responsibility of local and State governments.

Subsequent actions by these authorities are based on

projected or potential radiation exposures to

individuals in the population. The actions recommended

by these authorities have been identified as part of

their emergency planning efforts. The projected or

potential exposures are based on predictions made by

VEGP from parameters such as status of reactor core and

core cooling systems, effluent release rates, effluent

radioactivity levels, containment radiation levels, containment pressure, actual meteorological conditions, or measurements of offsite levels of radioactivity made

by VEGP and/or State radiological response teams.

The described emergency classes and the emergency action

levels which determine them are agreed on by VEGP and

State and local authorities. The emergency action

levels will be reviewed by these parties annually.

D.2 EMERGENCY CLASS DESCRIPTION AND RESPONSE EMERGENCY CLASS

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

• Notification of Unusual Event (NUE).

D-2 REV 48 03/08

• Alert.

• Site Area Emergency.

• General Emergency.

The NUE and Alert classes give early notification of

minor events that could lead to more serious

consequences or that might indicate more serious

conditions which have not yet fully developed. A Site

Area Emergency reflects conditions where some

significant releases are likely or are occurring but

where core melt is not likely to occur. In this

situation, onsite and offsite VEGP emergency personnel

and monitoring teams are fully mobilized. A General

Emergency involves actual or imminent substantial core

degradation or melting with the potential for loss of

containment. The immediate action for this class is to

recommend evacuation.

Emergency class descriptions are provided below. Also

included are the actions to be carried out by VEGP.

D.2.1 NOTIFICATION OF UNUSUAL EVENT

1. CLASS DESCRIPTION Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are

expected unless further degradation of safety

systems occurs.

2. VEGP ACTIONS

a. Inform State and local offsite authorities of the nature of the unusual event within 15

min. of classifying the emergency. Notify

the Nuclear Regulatory Commission (NRC) as

soon as possible (ASAP) but no later than 1

h following classification of the emergency.

b. Augment on-shift resources, as needed.

c. Assess and respond to the event.

d. Escalate to a more severe class, if appropriate, or close out with a verbal summary to offsite authorities followed by a

written summary within 24 h.

D-3 REV 48 03/08 D.2.2 ALERT

1. CLASS DESCRIPTION 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 U.S. Environmental Protection

Agency (EPA) Protective Action Guideline (PAG)

exposure levels.

2. VEGP ACTIONS

a. Within 15 min. of classification, inform State and local offsite authorities of

Alert Emergency and reasons for emergency.

Notify the NRC ASAP but no later than 1 h

following classification of the emergency.

b. Augment resources and activate the emergency response facilities (e.g.

Technical Support Center (TSC), Operational

Support Center (OSC), and the emergency

operating facility (EOF). These actions

may be delayed for security based events at

the discretion of the emergency director.

c. Assess and respond to the emergency.

d. Mobilize, and dispatch if necessary, onsite survey teams.

e. Provide periodic plant status updates to offsite authorities.

f. Provide periodic meteorological assessments to offsite authorities and, if any emergency releases are occurring, field

monitoring team readings or dose estimates

for actual releases.

g. Activate the Emergency Response Data System for the affected unit within 1 h following

declaration of the Alert.

h. Escalate to a more severe class, if appropriate, or close out emergency class by verbal summary to offsite authorities

followed by written summary within 8 h of

closeout.

D-4 REV 48 03/08 D.2.3 Site Area Emergency

1. CLASS DESCRIPTION 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 exceed EPA PAG exposure

levels except near the site boundary.

2. VEGP ACTIONS

a. Within 15 min. of classification, inform State and local offsite authorities of Site

Area Emergency and reasons for emergency.

Notify the NRC ASAP but no later than 1 h

following classification of the emergency.

b. If necessary, provide protective action recommendations to State and local

authorities.

c. Augment resources and activate the emergency response facilities (e.g.

Technical Support Center (TSC), Operational

Support Center (OSC), and the Emergency

Operating Facility (EOF. These actions may

be delayed for security based events at the

discretion of the emergency director.

d. Assess and respond to the emergency.

e. Dispatch as necessary onsite and offsite survey teams.

f. Dedicate individuals for plant status updates to offsite authorities and periodic press briefings.

g. On a periodic basis, make senior technical and management staff available for

consultation with the NRC and State

officials.

h. Provide meteorological data and field monitoring team readings or dose estimates

to offsite authorities. i. Provide release and dose projections based on available plant condition

information and foreseeable contingencies.

D-5 REV 48 03/08

j. Activate the Emergency Response Data System for the affected unit within 1 h following

declaration of the Site Area Emergency.

k. Escalate to General Emergency, if appropriate, or close out emergency class by briefing of offsite authorities followed

by written summary within 8 h of closeout

or class reduction. D.2.4 General Emergency

1. CLASS DESCRIPTION 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. Release can be reasonably expected to exceed EPA PAG exposure levels

offsite for more than the immediate site area.

2. VEGP ACTIONS

a. Within 15 min. of classification, inform State and local offsite authorities of

General Emergency and reason for emergency.

Notify the NRC ASAP but no later than 1 h

following classification of the emergency.

b. Provide protective action recommendations to State and local authorities based upon plant

conditions and/or actual or projected

releases of radioactive material.

c. Augment resources and activate the emergency response facilities (e.g. Technical Support

Center (TSC), Operational Support Center (OSC), and the Emergency Operating Facility (EOF)). These actions may be delayed for

security based events at the discretion of

the emergency director.

d. Assess and respond to the emergency.

e. Dispatch onsite and offsite survey teams.

f. Dedicate an individual for plant status updates to offsite authorities and periodic press briefings.

g. On a periodic basis, make senior technical and management staff available for

consultation with the NRC and State

officials.

h. Provide meteorological data and field monitoring team readings or dose estimates

to offsite authorities for actual releases.

D-6 REV 56 09/10

i. Provide release and dose projections based on plant condition and foreseeable

contingencies.

j. Activate the Emergency Response Data System for the affected unit within 1 h following

declaration of the General Emergency.

k. Close out emergency class by briefing of offsite authorities followed by written

summary within 8 h of closeout or class

reduction.

D.3 CLASSIFICATION PROCESS The procedure by which the plant operational staff classifies emergencies is NMP-EP-110, "Emergency Classification Determination and Initial Actions." This procedure is used to classify the emergency condition

upon recognition of an off-normal condition as

determined by: direction from another procedure;

instrument readings/alarms; a critical safety function

status on the safety parameters display system; direct

observation; or reported events. Plant procedures which

direct reference to this procedure include control room

annunciator response procedures, abnormal and emergency

operating procedures that reflect directly on Emergency

Action Levels (EAL), as well as chemistry and health

physics procedures that are used in the conduct of shift

activities. The Critical Safety Functions Status Trees (CSFSTs) contained in Emergency Operating Procedure

19200-C are an integral part of the EAL scheme and form

the bases for numerous EALs. The instrument readings, alarms, conditions, and event(s) which may result in an

emergency classification are as specified

in the EALs themselves or in the off-normal procedures

directing the operator to the Emergency Plan

implementing procedures.

EMERGENCY ACTION LEVEL (EAL):

A pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given

emergency class. An EAL can be: an instrument reading;

an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, observable event;

results of analyses; entry into specific emergency

operating procedures; or another phenomenon which, if it

occurs, indicates entry into a particular emergency

class. INITIATING CONDITION (IC)

An emergency condition which sets it apart from the broad class of conditions that may or may not have the potential to escalate into a

radiological emergency. It can be a continuous, D-7 REV 48 03/08 measurable function that is outside technical specifications, such as elevated RCS temperature or falling reactor coolant level (a symptom). It also encompasses occurrences such as FIRE (an event) or reactor coolant pipe failure (an event or a barrier breach). Recognition Categories ICs and EALs can be grouped in one of several schemes. This generic classification scheme incorporates symptom-based, event-based, and barrier-based ICs and EALs.

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

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

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

Status Tree Monitoring (or their equivalent). Challenge to one or more barriers generally is initially identified through instrument readings and periodic sampling. Under present barrier-based EALs, deterioration of the reactor coolant system pressure boundary or the fuel clad barrier usually indicates an "Alert" condition, two barriers under D-8 REV 48 03/08 challenge a Site Area Emergency, and loss of two barriers with the third barrier under challenge is a General Emergency. The fission product barrier matrix described in Section 5-F is a hybrid approach that recognizes that some events may represent a challenge to more than one barrier, and that the containment barrier is weighted less than the reactor coolant system pressure boundary and the fuel clad barriers.

Symptom-based ICs and EALs are most easily identified when the plant is in a normal startup, operating or hot shutdown mode of operation, with all of the barriers in place and the plant's instrumentation and emergency safeguards features fully operational as required by technical specifications.

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

Barrier-based ICs and EALs also are heavily dependent on the ability to monitor instruments that indicate the condition of plant operating and safety systems. Fuel cladding integrity and reactor coolant levels can be monitored through several indicators when the plant is in a normal operating mode, but this capability is much more limited when the plant is in a refueling mode, when many of these indicators are disconnected or off-scale. The need for this instrumentation is lessened, however, and alternate instrumentation is placed in service when the plant is shut down. It is important to note that in some operating modes there may not be definitive and unambiguous indicators of containment integrity available to control room personnel.

For this reason, barrier-based EALs should not place undue reliance on assessments of containment integrity in all operating modes. Generally, Technical Specifications relax maintaining containment integrity requirements in modes 5 and 6 in order to provide flexibility in performance of specific tasks during shutdown conditions. Containment pressure and temperature indications may not increase if there is a pre-existing breach of containment integrity. At D-9 REV 48 03/08 most plants, a large portion of the containment's exterior cannot be monitored for leakage by radiation monitors.

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

In many cases, a combination of symptom-, event- and barrier-based ICs will be present as an emergency develops.

In a loss of coolant accident (LOCA), for example:

• Coolant level is lowering; (symptom)

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

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

• At some level, fuel failure begins with indicators such as high off-gas, high coolant activity samples, etc. (barrier breach or symptom)

There are three considerations related to emergency classes.

These are:

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

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

Threshold Values The most common bases for establishing these boundaries are the technical specifications and setpoints for each plant D-10 REV 48 03/08 that have been developed in the design basis calculations and the Final Safety Analysis Report (FSAR).

For those conditions that are easily measurable and instrumented, the boundary is likely to be the EAL (observable by plant staff, instrument reading, alarm setpoint, etc.) that indicates entry into a particular emergency class. For example, the main steam line radiation monitor may detect high radiation that triggers an alarm.

That radiation level also may be the setpoint that closes the main steam isolation valves (MSIV) and initiates the reactor scram. This same radiation level threshold, depending on plant-specific parameters, also may be the appropriate EAL for a direct entry into an emergency class.

In addition to the continuously measurable indicators, such as coolant temperature, coolant levels, leak rates, containment pressure, etc., the FSAR provides indications of the consequences associated with design basis events.

Examples would include steam pipe breaks, MSIV malfunctions, and other anticipated events that, upon occurrence, place the plant immediately into an emergency class.

Another approach for defining these boundaries is the use of a plant-specific probabilistic safety assessment (PSA - also known as probabilistic risk analysis, PRA). PSAs have been completed for all individual plants. PSAs can be used as a good first approximation of the relevant ICs and risk associated with emergency conditions for existing plants.

Each plant has an Individual Plant Evaluation (IPE) and an Individual Plant Evaluation for External Events (IPEEE).

Generic insights from a PSA/PRA, the IPE, IPEEE and related severe accident assessments which apply to EALs and emergency class determinations are:

1. Core damage frequency at many BWRs is dominated by sequences involving prolonged loss of all AC power. In addition, prolonged loss of all AC power events are extremely important at PWRs. This would indicate that should this occur, and AC power is not restored within 15 minutes, entry into the emergency class at no lower than a Site Area Emergency, when the plant was initially at power, would be appropriate. This implies that precursors to loss of all AC power events should appropriately be included in the EAL structure.

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

D-11 REV 48 03/08

3. PRAs show that leading contributors to latent fatalities were containment bypass, large LOCA with early containment failure, station blackout greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (e.g., LOCA consequences of Station Blackout), and reactor coolant pump seal failure. This indicates that generic EAL methodology must be sufficiently rigorous to cover these sequences in a timely fashion.

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

Establishing EALs for such conditions must take estimated evacuation time into consideration to minimize the potential for the plume to pass while evacuation is underway.

Regardless of whether or not containment integrity is challenged, it is possible for significant radioactive inventory within containment to result in EPA PAG plume exposure levels being exceeded even assuming containment is within technical specification allowable leakage rates. With or without containment challenge, however, a major release of radioactivity requiring offsite protection actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.

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

D-12 REV 48 03/08 For discrete (discontinuous) events, the approach will have to be somewhat different. Typically, in this category are internal and external hazards such as FIRE or earthquake.

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

Plant emergency operating procedures (EOPs) are designed to maintain and/or restore a set of CSFs which are listed in the order of priority for restoration efforts during accident conditions. While the actual nomenclature of the CSFs may vary among plants, generally the PWR CSF set includes:

Subcriticality Core cooling Heat sink Pressure-temperature-stress (RCS integrity) Containment RCS inventory There are diverse and redundant plant systems to support each CSF. By monitoring the CSFs instead of the individual system component status, the impact of multiple events is inherently addressed, e.g., the number of operable components available to maintain the critical safety function.

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

As an example, consider that the Westinghouse Owner's Group (WOG) Emergency Response Guidelines (ERGs) classify D-13 REV 48 03/08 challenges as YELLOW, ORANGE, and RED paths. If the core exit thermocouples exceed 1200 degrees F or 711 degrees F with low reactor vessel water level, a RED path condition exists. The ERG considers a RED path as "... an extreme challenge to a plant function necessary for the protection of the public ..." This is almost identical to the present NRC NUREG-0654 description of a site area emergency "...

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

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 thresholds has been exceeded. While this is particularly prudent at the higher emergency classes (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classes. Multiple Events and Emergency Class Upgrading The SNC Classification procedures are written to classify events based on meeting the Initiating Condition (IC) and a Threshold Value (TV) for an EAL considering each Unit independently. Two IC Matrices are used, one for Hot ICs and one for Cold ICs. The temperature criteria of the Cold Shutdown Mode determines if the unit should use the Hot or Cold Matrix.

The IC Matrices are human factored to read from top to bottom General Emergency to Notification of Unusual Event within a category or subcategory to eliminate the higher classifications before reaching a lower classification.

This arrangement lessens the possibility of under-classifying a condition.

During events, the ICs and TVs are monitored and if conditions meet another higher EAL, that higher emergency classification is declared and appropriate notifications made. Notifications are made on a site basis. If both units are in concurrent classifications, the highest classification would be used for the notification and the other unit classification noted on the notification form.

There are six EALs which specifically state that if the condition cannot be mitigated and is imminent, the Emergency D-14 REV 57 06/12 Director should not wait for the time to elapse and should

classify immediately (RG1, RS1, RA1, RU1, CA4, CU4). These

six EALs are conditions which indicate a loss of control of

radioactive materials, or could lead to a significant

radioactive release or core damage situation and therefore, by making the anticipatory classification, onsite and

offsite protective actions associated with the emergency

classification level can then be initiated as quickly as

possible.

At all times, when conditions present themselves that are

not explicitly provided in the EAL scheme the Emergency

Director has discretion to declare an event based on his

knowledge of the emergency classes and judgment of the

situation or condition. Specific EALs (HU5, HA6, HS3, HG2)

are provided within the scheme to allow these discretionary

classifications.

Classification Timeliness The emergency declaration process starts with information being available to plant operators to recognize an off-normal plant condition via indications on plant instrumentation including alarms, or via reports from other plant personnel (e.g., reports of fire) or from persons outside of the plant (e.g., severe weather warnings). The plant operators assess the validity of these indications or reports by checking instruments, comparing indications on redundant instruments, or dispatching personnel to confirm reports. After validating the indication or report, the plant operators then compare the off-normal condition to the EAL thresholds in the emergency classification scheme. Not all off-normal conditions are immediately obvious, and not all indications are unambiguous. While some conditions can be classified upon recognition, others require further assessment.

The capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an EAL has been exceeded has been established and is outlined in applicable procedures. Emergency conditions are classified promptly upon identification that an emergency action leve3l (EAL) threshold has been exceeded.

The 15-minute period encompasses all assessment, classification, and declaration actions associated with making an emergency declaration form the first availability of a plant indication or receipt of a report of an off-normal condition by plant operators up to and including the declaration of the emergency. If classifications and declarations are performed away from the CR, all delays incurred in transferring information from the CR (where the alarms, indications, and reports are first received) to the D-15 REV 57 06/12 ERF (at which declarations are made) are included within the 15-minute criterion.

Classification Downgrading

The SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower classification.

Termination criteria contained in the Emergency Plan

Implementing Procedures shall be completed for an event to

be terminated. At termination, on an event specific basis, the site can either enter normal operating conditions or

enter a recovery condition with a recovery organization

established for turnover from the ERO.

Transient Events

The NEI 99-01 EAL development philosophy stated in Section

3.6 "-The approach taken in the development of these EALs is

based on risk assessment to set the boundaries of the

emergency classes and assure that all EALs that trigger that

emergency class are in the same range of relative risk.

Precursor conditions of more serious emergencies also

represent a potential risk to the public and must be

appropriately classified."

Using this philosophy, if an IC and TV are met, the threat

associated with the emergency class is present and should

therefore be classified. If a condition existed for a short

period of time, but is no longer present, the associated

threat is no longer applicable. Therefore, classification

of these events is not appropriate in that the actions taken

associated with the threat level are no longer needed. To

that end, the condition should be reported using the

guidance of NUREG 1022, Rev 1, Section 3.1.1 which states in

part:

"Occasionally, a licensee may discover that an event or

condition had existed which met the emergency plan criteria

but that no emergency had been declared and the basis for

the emergency class no longer exists at the time of this

discovery. This may be due to a rapidly concluded event or

an oversight in the emergency classification made during the

event or it may be determined during a post-event review.

Frequently, in cases of this nature, which were discovered

after the fact, licensees have declared the emergency class, immediately terminated the emergency class and then made the

appropriate notifications. However, the staff does not

consider actual declaration of the emergency class to be

necessary in these circumstances; an ENS notification (or an

ENS update if the event was previously reported but

misclassified) within one hour of the discovery of the

undeclared (or misclassified) event will provide an

acceptable alternative. Notification of the State and local

emergency response organizations should be made in D-16 REV 57 06/12 accordance with the arrangements made between the licensee

and offsite organizations."

Many of the Threshold Values associated with EALs are based

on a time period for their condition to be met. Generally

these are 15 minutes, but 30 and 60 minutes are also used.

The time provided in these Threshold Values is to allow

actions to be taken to reduce or eliminate the threat

presented by the condition. With time criteria associated

with the Threshold Values, most of the transient conditions

have been addressed and bounded within the EALs.

D.4 DEFINITIONS D.4.1 HOSTILE ACTION

An act toward a nuclear power plant (NPP) or its

personnel that includes the use of violent force to

destroy equipment, takes 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 the NPP. Non-

terrorism-based EALs should be used to address such

activities, (e.g., violent acts between individuals in

the owner controlled area.)

D.4.2 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.

D-17 REV 62 10/14 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Any UNPLANNED Release of Gaseous or Li quid Radioactivity to the Environment that Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or Longer.

Operating Mode Applicability:

All Threshold Values:

(1 OR 2) Note: The Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. 1. VALID reading on any of the following effluent monitor that exceeds two times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes OR longer. Liquid Radwaste Effluent Line (RE-0018)

SG Blowdown Effluent Line (RE-0021)

Turbine Bldg Effluent Line (RE-0848)

Gaseous Radwaste (ARE-0014)

Turbine Bldg Vent, SJAE (RE-12839)

Plant Vent (RE-12442C or RE-12444C)

2. Confirmed sample analyses for gaseous OR liquid releases indicates concentrations OR release rates, with a release duration of 60 minutes OR longer, in excess of two times Technical Specification 5.5.4, as confirmed by the ODCM.

Basis: UNPLANNED, as used in this context, 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. The Emergency

Director should not wait until 60 minutes has elapsed, but should declare the event as soon as D-18 REV 57 06/12 it is determined that the release duration has or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency

Director should, in the absence of data to the contrary, assume that the release has exceeded

60 minutes.

This IC addresses a potential or actual decline 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 adm inistrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in

the Offsite Dose Calculation Manual (ODCM). The occurrence of extended, uncontrolled

radioactive releases to the environment is indicative of a degradation in these features and/or

controls.

Threshold Value #1 addresses radioactivity releases, that for whatever reason, cause effluent

radiation monitor readings to exceed two times the Technical Specification limit and releases

are not terminated within 60 minutes. This alarm setpoint may be associated with a planned

batch release, or a continuous release path. In either case, the setpoint is established by the

ODCM to warn of a release that is not in compliance with the TS 5.5.4. Indexing the Threshold

Value to the ODCM setpoints in this manner ens ures that the Threshold Value will never be less than the setpoint established by a specific discharge permit.

Threshold Value #2 is intended for effluent monitoring on non-routine release pathways for

which a discharge permit would not normally be prepared. The ODCM establishes a

methodology for determining effluent radiation monitor setpoints. The ODCM specifies default

source terms and, for gaseous releases, prescribes the use of pre-determined annual average

meteorology in the most limiting downwind sector for showing compliance with the regulatory

commitments. These monitor reading Threshold Values are determined using this methodology.

Threshold Value #3 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.

D-19 REV 57 06/12 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Unexpected Rise in Plant Radiation.

Operating Mode Applicability

All Threshold Values: (1 OR 2) 1. a. VALID indication of uncontrolled water level lowering in the reactor refueling cavity, as indicated by spent fuel pool, OR fuel transfer canal with all irradiated fuel assemblies remaining covered by water by any of the following:

Personnel report of low water level LSHL-0625 offscale low (ALB05 E02)

Personnel report of cavitation OR low discharge pressure for SFP (1/2-1213-P6-002 OR -005) AND/OR RHR (1/2-1205-P6-001 OR -002) Pumps AND b. UNPLANNED VALID Direct Area Radiation Monitor readings rise on any of the following:

RE-0008 in the fuel building RE-0002, -0003, -0004 in containment RE-0011 at the seal table RE-0005, -0006 in containment

2. UNPLANNED VALID Direct Area Radiation Monitor readings rise by a factor of 1000 over normal* levels.

• Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution

and requires corrective or mitigative actions.

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 D-20 REV 57 06/12 condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses raised radiation levels as a result of water level lowering above the RPV

flange or events that have resulted, or may result, in unexpected rises in radiation dose rates

within plant buildings. These radiation rises represent a loss of control over radioactive material

and may represent a potential degradation in the level of safety of the plant.

In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the

Spent Fuel Pit/Fuel Transfer Canal at a BWR, explicit coverage of these types of events via

Threshold Value #1 is appropriate given their potential for raised doses to plant staff.

Classification as a NOUE is warranted as a precursor to a more serious event. Indications

include instrumentation such as water level and local area radiation monitors, equipment

parameters and personnel (e.g., refueling crew) reports. If available, video cameras may allow

remote observation. Depending on available level instrumentation, the declaration threshold

may need to be based on indications of water makeup rate or lowering in refueling water

storage tank level.

Threshold Value 1a primary indicator is visual observation.

Threshold Value 1b limits: The dose above the fuel in the core or storage racks is 1.27 x 10 4 mR/hr with eight feet of water shielding and a reduction of 1.27 x 10 4 / 6.38 x 10 2 20 per two feet of water. . These monitors do not directly see the fuel, but see the dose reflected from the

fuel building ceiling or containment dome. A conservative reduction on reflection is 1.14 x 10

-2.

For the containment monitors, the distance from the approximate midline of the dome to the operating deck is 140 ft, so the just covered dose for RE-0005 and -0006 would be

1.8 x 10 6 R/hr x 0.0114 x (13.7/2)2 ft 2 / 140 2 ft 2 162 R/hr.

To remain on-scale, RE-0002, -0003 (range of 5.4 x 10 3 mR/hr,) equivalent reading at the pool surface would be 5.4 x 10 3 x (1.8 x 10 6/162) 6.0 x 10 4 R/hr. This requires a dose reduction of 1.8 x 10 6 / 6.0 x 10 4 30 or 24 x (ln 30 / ln 20) 27" water cover, and RE-0004 (range of 1 x 10 4 mR/hr,) requires about 22" of water cover.

For the seal table monitor on level A, RE-0011, the operating deck at elevation 220' is 2.75 feet

thick, which is approximately equivalent to 2.75 ft x 2.16 g/cc / 1.0 g/cc = 6 ft of water or a dose

reduction factor of 20 6/2 = 8 x 10 3 yielding a dose of:

1.8 x 10 6 R/hr x 0.0114 x (13.7/2)2 ft 2 / (140+20) 2 ft 2 / 8 x 10 3 15 mR/hr.

In the fuel handling building, the rack dose will be reflected from the ceiling at elevation 261'-8"

or about 40 ft from the floor yielding a just covered dose rate of

1.8 x 10 6 R/hr x 0.0114 x (13.7/2)2 ft 2 / (40)2 ft 2 1.9 x 10 3 R /hr. This will require a dose reduction of 1.9 x 10 6/ 1 x 10 4 189 or 24 x (ln 189 / ln 20) 42" water cover to stay on scale.

The dose at RE-0008 would be D-21 REV 57 06/12 1.45 x 2.5 mR/hr x 20 10/2 x 0.0114 x (6.5)2 ft 2 / 40 2 ft 2 = 1.1 x 10 4 mR/hr. To remain on-scale, RE-0008 requires 24 x ln (1.1 x 10 4 / 1.0 x 10

4) / ln 20 1" water.

While other radiation monitors could detect a rise 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. For example, the reading

on an area radiation monitor located on the refueling bridge may raise due to planned

evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast.

Generally, raised 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. This event escalates to an Alert per IC AA2 if

irradiated fuel outside the reactor vessel is uncovered. For events involving irradiated fuel in the

reactor vessel, escalation would be via the Fission Product Barrier Matrix for events in operating

modes 1-4.

Threshold Value #2 addresses UNPLANNED rises in in-plant radiation levels that represent a

degradation in the control of radioactive material, and represent a potential degradation in the

level of safety of the plant. This event escalates to an Alert per IC AA3 if the increase in dose

rates impedes personnel access necessary for safe operation.

D-22 REV 62 10/14 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RA1 Initiating Condition -- ALERT Any UNPLANNED Release of Gaseous or Li quid Radioactivity to the Environment that Exceeds 200 Times the Radiological E ffluent Technical Specifications for 15 Minutes or Longer.

Operating Mode Applicability:

All Threshold Values:

(1 OR 2) Note: The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

1. VALID reading on any of the following effluent monitor that exceeds 200 times the alarm setpoint established by a current radioactivity discharge permit for 15 minutes OR longer. Monitor Liquid Radwaste Effluent Line (RE-0018)

SG Blowdown Effluent Line (RE-0021)

Turbine Bldg Effluent Line (RE-0848)

Gaseous Radwaste (ARE-0014)

Turbine Bldg Vent, SJAE (RE-12839)D-22 Plant Vent (RE-12444C or RE-12442C)

2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations OR release rates in excess of 200 times Technical Specification 5.5.4 as confirmed by the

ODCM , with a release duration of 15 minutes OR longer. Basis:

D-23 REV 62 10/14 UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an

instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct

observation by plant personnel, such that doubt related to the indicator's operability, the

condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses a potential or actual decline 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 adm inistrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in

the Offsite Dose Calculation Manual (ODCM). The occurrence of extended, uncontrolled

radioactive releases to the environment is indicative of a degradation in these features and/or

controls.

The Radiological Effluent Control Plan (RECP) multiples are specified in ICs RU1 and RA1 only

to distinguish between non-emergency conditions, and from each other. While these multiples

obviously correspond to an offsite 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. Releases should not be prorated or averaged.

UNPLANNED, as used in this context, 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. The Emergency

Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it

is determined that the release duration has or will likely exceed 15 minutes. Also, if an ongoing

release is detected and the starting time for that release is unknown, the Emergency Director

should, in the absence of data to the contrary, assume that the release has exceeded 15

minutes.

Threshold Value #1 addresses radioactivity releases that for whatever reason cause effluent

radiation monitor readings that exceed two hundred times the alarm setpoint established by the

radioactivity discharge permit. This alarm setpoint may be associated with a planned batch

release, or a continuous release path. In either case, the setpoint is established by the ODCM

to warn of a release that is not in compliance with the RECP. Indexing the Threshold Value

threshold to the ODCM setpoints in this manner ensures that the Threshold Value threshold will

never be less than the setpoint established by a specific discharge permit. Setpoints are 100

times those of RU1 or the maximum range of the monitor.

D-24 REV 62 10/14 Threshold Value #2 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.

Threshold Value #1 directly correlates with the IC since annual average meteorology is required to be used in showing compliance with the RECP and is used in calculating the alarm setpoints.

Thus, there will likely be a numerical inconsistency.

However, the fundamental basis of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant implied by

the uncontrolled release.

Due to the uncertainty associated with me teorology, emergency implementing procedures should call for the timely performance of dose assessments using actual (real-time) meteorology

in the event of a gaseous radioactivity release of this magnitude. The results of these

assessments should be compared to the ICs RS1 and RG1 to determine if the event

classification should be escalated.

D-25 REV 62 10/14 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RA2 Initiating Condition -- ALERT Damage to Irradiated Fuel OR Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.

Operating Mode Applicability

All Threshold Values: (1 OR 2) 1. UNPLANNED VALID alert alarm on any of the following radiation monitors:

Fuel Handling Building RE-008 CNMT BLDG Low Range** RE-002/003 **Mode 6 only during fuel transfer Fuel Handling BLDG EFFL. ARE-2532 A/B Fuel Handling BLDG EFFL. ARE-2533 A/B

2. Loss of water level that has or will result in the uncovering of irradiated fuel outside the Reactor Vessel as indicated by ANY of the following:

Personnel report during fuel assembly movements.

Spent Fuel Pool Storage Less than El 193'-5" Transfer Canal Transit Elevation Less than El 187'-2" Reactor Core Elevation Less than El 181'-10" (63% on Full Range RVLIS)

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution

and requires corrective or mitigative actions.

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

This IC addresses specific events that have resulted, or may result, in unexpected rises 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 a degradation in the level of safety of the plant. These events escalate from IC RU2 in that fuel activity has been released, or is anticipated due to fuel heatup. This IC applies to D-26 REV 57 06/12 spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

Threshold Value #1 addresses radiation monitor indications of fuel uncovery and/or fuel damage. Raised readings on ventilation monitors may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Raised background at the monitor due to water level lowering may mask raised ventilation exhaust airborne activity and needs to be considered. While a radiation monitor could detect a rise 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. For example, the 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. Application of these Initiating Conditions requires understanding of the actual radiological conditions present in the vicinity of the monitor. Information Notice No. 90-08, "KR-85 Hazards from Decayed Fuel" should be considered in establishing radiation monitor Threshold Values.

In Threshold Value #2, indications include water level and personnel reports. Visual observation

will be the primary indicator for spent fuel pool and fuel movement activities. Personnel report of personnel during fuel assembly movements is included to ensure that reports of actual or potential fuel uncovery is classified. If available, video cameras may allow remote observation.

Depending on available level indication, the declaration threshold may need to be based on

indications of water makeup rate or lowering in refueling water storage tank level.

D-27 REV 57 06/12 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RA3 Initiating Condition -- ALERT Release of Radioactive Material or Rises in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown

Operating Mode Applicability

All Threshold Values: (1 OR 2) 1. VALID radiation monitor readings greater than 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions:

Control Room radiation monitor RE-001 Central Alarm Station (by survey)

2. VALID radiation readings greater than 1 R/hr values in areas requiring infrequent access to maintain plant safety functions.

Turbine Building South & North Main Steam Valve Rooms Control Building Fire Pump House Aux Feed Water Pump House Low Voltage Switchyard Diesel Generator Buildings NSCW Towers Basis: VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses raised radiation levels that impede necessary access to operating stations, or

other areas containing equipment that must be operated manually or that requires local

monitoring, in order to maintain safe operation or perform a safe shutdown. It is this impaired

ability to operate the plant that results in the actual or potential substantial degradation of the

level of safety of the plant. The cause and/or magnitude of the rise in radiation levels is not a

concern of this IC. The Emergency Director must consider the source or cause of the raised

radiation levels and determine if any other IC may be involved. For example, a dose rate of 15

mR/hr in the control room may be a problem in itself. However, the rise may also be indicative of

high dose rates in the containment due to a LOCA. In this latter case, an SAE or GE may be

indicated by the fission product barrier matrix ICs.

D-28 REV 57 06/12 At multiple-unit sites, the Threshold Values could result in declaration of an Alert at one unit due to a radioactivity release or radiation shine resulting from a major accident at the other unit. This

is appropriate if the rise impairs operations at the operating unit.

This IC is not meant to apply to raises in the containment dome radiation monitors as these are events which are addressed in the fission product barrier matrix ICs. Nor is it intended to apply to anticipated temporary rises due to planned events (e.g., incore detector movement, radwaste

container movement, depleted resin transfers, etc.)

Areas requiring continuous occupancy includes the control room and the central alarm station.

The value of 15mR/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 that the 15 mR/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.

For areas requiring infrequent access, the 1 R/hr (Locked High Rad Area) is be based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e., 10 CFR 20), and in doing so, will

impede necessary access. As used here, impede, includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant.

D-29 REV 62 10/14 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RS1 Initiating Condition -- SITE AREA EMERGENCY Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR TEDE OR 500 mR Thyroid CDE for the Actual or Projected Duration of the Release.

Operating Mode Applicability:

All Threshold Values:

(1 OR 2 OR 3) Note: If dose assessment results are available at the time of declaration, the classification should be based on Threshold Value #2 instead of Threshold Value #1.While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently

escalated.

Note: The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

1. VALID reading on any of the following radiation monitors that exceeds OR is expected to exceed the reading shown for 15 minutes OR longer:

RE-12839E 5 Ci/cc RE-12444E 1 x 10 2 Ci/cc 2. Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE OR 500 mR thyroid CDE at OR beyond the site boundary.

3. Field survey results indicate closed window dose rates exceeding 100 mR/hr expected to continue for more than one hour; OR analyses of field survey samples indicate thyroid CDE of 500 mR for one hour of inhalation, at OR beyond the site boundary.

Basis: 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 D-30 REV 57 06/12 condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses radioactivity releases that result in doses at or beyond the site boundary that

exceed a small fraction 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. While these failures are addressed by other ICs, this IC provides appropriate diversity

and addresses events which may not be able to be classified on the basis of plant status alone, e.g., fuel handling accident in spent fuel building.

The TEDE dose is set at 10% of the EPA PAG, while the 500 mR thyroid CDE was established

in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

The Emergency Director should not wait until 15 minutes has elapsed, but should declare the

event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

The site specific monitor list in Threshold Value #1 includes monitors on all potential release

pathways.

The monitor reading Threshold Values are determined using a dose assessment method that

back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used for determining the

monitor reading Threshold Values in ICs RU1 and RA1. This protocol maintains intervals

between the Threshold Values for the four classifications. Since doses are generally not

monitored in real-time, a release duration of one hour is assumed, and that the Threshold

Values be based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500

mR/hour thyroid, whichever is more limiting (as was done for Threshold Value #3).

The release rates which result in site boundary doses of 100 mR TEDE are in excess of the

range of the monitors listed in RU1 and RA1.

Since dose assessment is based on actual meteorology, whereas the monitor reading

Threshold Values are not, the results from these assessments may indicate that the

classification is not warranted, or may indicate that a higher classification is warranted. For this

reason, emergency implementing procedures shoul d call for the timely performance of dose assessments using actual meteorology and release information. If the 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 Threshold

Values. Classification should not be delayed pending the results of these dose assessments.

D-31 REV 64 3/15 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT RG1 Initiating Condition -- GENERAL EMERGENCY Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mR TEDE OR 5000 mR Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology.

Operating Mode Applicability:

All Threshold Values:

(1 OR 2 OR 3) Note: If dose assessment results are available at the time of declaration, the classification should be based on Threshold Value #2 instead of Threshold Value #1.While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently

escalated.

Note: The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

1. VALID reading on any of the following radiation monitors that exceeds OR expected to exceed the reading shown for 15 minutes OR longer:

RE-12839E 50 Ci/cc RE-12444E 2.1 x 10 3 Ci/cc

2. Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE OR 5000 mR thyroid CDE at OR beyond the site boundary.

3. Field survey results indicate closed window dose rates exceeding 1000 mR/hr expected to continue for more than one hour; OR analyses of field survey samples indicate thyroid CDE of 5000 mR for one hour of inhalation, at OR beyond site boundary.

Basis: 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 D-32 REV 57 06/12 condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC 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. While these failures are addressed by

other ICs, this IC provides appropriate diversity and addresses events which may not be able to

be classified on the basis of plant status alone. It is important to note that, for the more severe

accidents, the release may be unmonitored or there may be large uncertainties associated with

the source term and/or meteorology.

The Emergency Director should not wait until 15 minutes has elapsed, but should declare the

event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

The site specific monitor list in Threshold Value #1 includes monitors on all potential release

pathways.

The monitor reading Threshold Values are determined using a dose assessment method that

back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used for determining the

monitor reading Threshold Values in ICs RU1 and RA1. This protocol will maintain intervals

between the Threshold Values for the four classifications. Since doses are generally not

monitored in real-time, a release duration of one hour is assumed, and that the Threshold

Values are based on a site boundary (or beyond) dose of 1000 mR/hour whole body or 5000

mR/hour thyroid, whichever is more limiting.

Since dose assessment is based on actual meteorology, whereas the monitor reading

Threshold Values are not, the results from these assessments may indicate that 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 the 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 Threshold

Values. Classification should not be delayed pending the results of these dose assessments.

D-33 REV 48 03/08 SYSTEM MALFUNCTION CU1 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT RCS Leakage.

Operating Mode Applicability:

Cold Shutdown Threshold Values:

1. Unable to establish or maintain pressurizer level greater than 17%.

Basis: This IC is included as a NOUE because it is considered to be a potential degradation of the level of safety of the plant. The inability to establish and maintain level is indicative of loss of RCS inventory. Prolonged loss of RCS Inventory may result in escalation to the Alert level via either IC CA1 (Loss of RCS) or CA4 (Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV).

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and RCS inventory and level monitoring means such as Pressurizer level indication and makeup volume control tank levels are normally available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

D-34 REV 56 09/10 SYSTEM MALFUNCTION CU2 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of RCS Inventory wit h Irradiated Fuel in the RPV.

Operating Mode Applicability:

Refueling Threshold Values:

(1 OR 2) 1. UNPLANNED RCS level lowering below 194' (RPV flange) for greater than or equal to 15 minutes 2. a. RPV level CANNOT be monitored AND b. A loss of RCS inventory may be occurring as indicated by unexplained level rise in ANY of the following:

Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT)

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution

and requires corrective or mitigative actions.

This IC is included as a NOUE 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. Refueling evolutions that lower RCS 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 warrants declaration of a NOUE 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 any of the redundant means of refill that should be available. If level cannot be restored in this time frame then it may indicate a more serious condition exists. Continued loss of RCS Inventory will result in escalation to the Alert level via either IC CA2 (Loss of RPV Inventory with Irradiated Fuel in the RPV) or CA4 (Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV).

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling modes. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not

intact and RPV level and inventory are monitored by different means.

In the refueling mode, normal means of core temperature indication and RCS level indication may not be available. Redundant means of RPV level indication will normally be installed D-35 REV 56 09/10 (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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 Alert would be via either CA2 or RCS heatup

via CA4.

Threshold Value 1 involves a lowering in RCS level below the top of the RPV flange that continues for 15 minutes due to an UNPLANNED event. This Threshold Value is not applicable to lowering levels in flooded reactor cavity level (covered by RU2 Threshold Value1) until such time as the level lowers to the level of the vessel flange. Drawing AX4DR023 provides the RPV flange level at 194' 11/2" rounded to 194' for the threshold Value. If RPV level continues to lower and reaches the Bottom ID of the RCS Loop then escalation to CA2 would be appropriate. Note that the Bottom ID of the RCS Loop Setpoint should be the level equal to the bottom of the RPV

loop penetration (not the low point of the loop).

D-36 REV 48 03/08 SYSTEM MALFUNCTION CU3 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability:

Cold Shutdown Refueling Threshold Value:

1. a. Loss of power to OR from transformers 1(2)NXRA AND 1(2)NXRB resulting in loss of all off-site electrical power to BOTH 1(2)AA02 AND 1(2)BA03 for greater than 15 minutes.

AND b. At least one emergency diesel generator supplying power to EITHER AA02 OR BA03. 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.

Capability to cross-tie AC power from a companion unit may take credit for the redundant power source in the associated Threshold Value for this IC. Inability to effect the cross-tie within 15 minutes warrants declaring a NOUE.

D-37 REV 48 03/08 SYSTEM MALFUNCTION CU4 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV. Operating Mode Applicability:

Cold Shutdown Refueling Threshold Values:

(1 OR 2) Note: The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the duration has or will likely exceed the Threshold Value.

1. An UNPLANNED event results in RCS temperature exceeding 200

°F. 2. Loss of all RCS temperature AND RPV level indication for greater than 15 minutes.

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC is included as a NOUE 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 RCS 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 typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. 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 the 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 CA4 or CA1 will occur if required.

During refueling the level in the RPV will normally be maintained above the RPV flange.

Refueling evolutions that lower water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid rises in RCS/RPV temperatures depending on the time since shutdown. Escalation to the Alert level via CA4 is provided should an UNPLANNED event result in RCS temperature exceeding the Technical Specification cold s hutdown temperature limit with CONTAINMENT CLOSURE not established.

D-38 REV 48 03/08 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 that 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 of refueling modes, Threshold Value 2 would result in declaration of a NOUE 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 via CA2 based on an inventory loss or CA4 based on exceeding its temperature criteria.

The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the Threshold Value 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.

D-39 REV 48 03/08 SYSTEM MALFUNCTION CU6 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of All Onsite OR Offsite Communications Capabilities.

Operating Mode Applicability:

Cold Shutdown Refueling Threshold Values:

(1 OR 2) 1. UNPLANNED loss of ALL of the following on-site communications capability affecting the ability to perform routine operations:

In plant telephones Public address system Plant radio systems

2. UNPLANNED loss of ALL of the following off-site communications capability:

ENN (Emergency Notification Network)

ENS (Emergency Notification System) Commercial phones (Radio, PBX, Satellite, Wireless)

VOIP (Voice Over Internet Protocol) OPX (Off Premise Extension)

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

The purpose of this IC and its associated Threshold Values 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 offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This Threshold Value is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

The list of onsite communications loss encompasses the loss of all means of routine communications.

D-40 REV 48 03/08 The list of offsite communications loss encompasses the loss of all means of communications with offsite authorities.

D-41 REV 48 03/08 SYSTEM MALFUNCTION CU7 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Required DC Power for Greater Than 15 Minutes.

Operating Mode Applicability:

Cold Shutdown Refueling Threshold Value:

1. a. UNPLANNED loss of Vital DC power to 125 VDC Buses AD1, BD1, CD1, AND DD1 indicated by bus voltage indications less than 105 VDC AND b. Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

The purpose of this IC and its associated Threshold Values 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 Threshold Value 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.

UNPLANNED is included in this IC and Threshold Value to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely plants will perform maintenance on a Train related basis during shutdown periods. 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 per CA4 "Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV." 105 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads.

D-42 REV 48 03/08 SYSTEM MALFUNCTION CU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent Criticality.

Operating Mode Applicability:

Cold Shutdown Refueling Threshold Values:

1. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC 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 and inadvertent dilution events. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC 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 IC SU8. The term "sustained" is used in order to allow exclusion of expected short term positive startup rates from planned fuel bundle or control rod movements during core alterations. These short term positive startup rates are the result of the rise in neutron population due to subcritical multiplication.

D-43 REV 62 10/14 SYSTEM MALFUNCTION CA1 Initiating Condition --

ALERT Loss of RCS Inventory.

Operating Mode Applicability:

Cold Shutdown Threshold Values:

(1 OR 2) 1. Loss of RCS inventory as indicated by RPV level less than elevation 185'-10" (73% on Full Range RVLIS)

2. a. RCS level CANNOT be monitored for greater than 15 minutes AND b. A loss of RCS inventory may be occurring as indicated by unexplained level rise in ANY of the following:

Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT)

Basis: These Threshold Values serve as precursors to a loss of ability to adequately cool the fuel. 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 lowering and potential core uncovery. This condition will result in a minimum classification of Alert. The Bottom ID of the RCS Loop Setpoint was chosen because at this level remote RCS level indication may be lost and loss of suction to decay heat removal systems has occurred. 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 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 or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. 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 the entry into cold shutdown was following a refueling). The above forms the basis

for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2)

In the cold shutdown mode, normal RCS leve l and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated D-44 REV 48 03/08 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 CS1 Site Area Emergency Threshold Value duration. The 15-minute duration allows CA1 to be an effective precursor to CS1. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CG1 basis. Therefore this Threshold Value meets the definition for an Alert.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

If RPV level continues to lower then escalation to Site Area will be via CS1 (Loss of Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV)

D-45 REV 62 10/14 SYSTEM MALFUNCTION CA2 Initiating Condition --

ALERT Loss of RPV Inventory with I rradiated Fuel in the RPV.

Operating Mode Applicability:

Refueling Threshold Values:

(1 OR 2) 1. Loss of inventory as indicated by RPV level less than 185'-10" (73% on Full Range RVLIS)

2. a. RPV level CANNOT be monitored for greater than 15 minutes AND b. A loss of RCS inventory may be occurring as indicated by unexplained level rise in ANY of the following:

Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT)

Basis: These Threshold Values serve as precursors 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 lowering and potential core uncovery. This condition will result in a minimum classification of Alert. The Bottom ID of the RCS Loop Setpoint was chosen because at this level remote RCS level indication may be lost and loss of suction to decay heat removal systems may occur. 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 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 or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />

{site-specific} or longer after the reactor has been shutdown. 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 the entry into cold shutdown was following a refueling). The above forms

the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).

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 that the ability to monitor level will not be interrupted. However, if all level D-46 REV 48 03/08 indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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 Site Area Emergency Threshold Value duration. The 15-minute duration allows CA2 to be an effective precursor to CS2. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CG1 basis. Therefore this Threshold Value meets the definition for an Alert.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

D-47 REV 48 03/08 SYSTEM MALFUNCTION CA3 Initiating Condition --

ALERT Loss of All Offsite Power AND Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability:

Cold Shutdown Refueling Defueled Threshold Value:

1. a. Loss of power to OR from transformers 1(2)NXRA AND 1(2)NXRB resulting in loss of all off-site electrical power to BOTH 1(2)AA02 AND 1(2)BA03.

AND b. Failure of emergency diesel generators to supply power to emergency busses.

AND c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.

Basis: Loss of all AC power compromises all plant sa fety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. 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 busses, relative to that specified for the Site Area Emergency Threshold Value. Escalating to Site Area Emergency IC SS1, if appropriate, is by Abnormal Rad Levels / Radiological Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable.

D-48 REV 48 03/08 SYSTEM MALFUNCTION CA4 Initiating Condition -- ALERT Inability to Maintain Plant in Cold S hutdown with Irradiated Fuel in the RPV.

Operating Mode Applicability:

Cold Shutdown Refueling Threshold Values:

(1 OR 2 OR 3) 1. An UNPLANNED event results in RCS temperature exceeding 200

°F with: a. CONTAINMENT CLOSURE NOT established AND b. RCS integrity NOT established Note 1 The Emergency Director should not wait until the indicated time of Threshold Values 2 or 3 has elapsed, but should declare the event as soon as it is determined that the duration has or will likely be exceeded

. Note 2 If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then Threshold Values 2 and 3 are not applicable.

2. An UNPLANNED event results in RCS temperature exceeding 200

°F for greater than 20 minutes (Note) with:

a. CONTAINMENT CLOSURE established AND b. RCS integrity NOT established OR c. RCS inventory reduced.

3. An UNPLANNED event results in:

a. RCS temperature exceeding 200

°F for greater than 60 minutes (Note)

OR b. RCS pressure increasing greater than 10 psig D-49 REV 48 03/08 Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling.

Threshold Value 1 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 (e.g., no freeze seals or nozzle dams). No delay time is allowed for Threshold Value1 because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment.

Threshold Value 2 addresses the complete loss of functions required for core cooling for GREATER THAN 20 minutes during refueling and cold shutdown modes when CONTAINMENT CLOSURE is established but RCS integrity is not established or RCS inventory is reduced (e.g., mid loop operation in PWRs). As in Threshold Value 1, 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 (e.g., no freeze seals or nozzle dams). The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible.

Threshold Value 3 addresses complete loss of functions required for core cooling for greater than 60 minutes during refueling and cold shutdown modes when RCS integrity is established. As in Threshold Value 1 and 2, 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 Threshold Value is immaterial given that 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 rise covers situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS pressure setpoint chosen is 10 psig and can be read on Control Board instrumentation. The Note indicates that Threshold Value 3 is not applicable if actions are successful in restoring an RCS heat removal system to operation and RCS temperature is being reduced within the 60 minute time frame assuming that the RCS pressure rise has remained less than the site specific pressure value.

A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary UNPLANNED excursion above 200

°F when the heat removal function is available.

The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the Threshold Value 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.

D-50 REV 62 10/14 SYSTEM MALFUNCTION CS1 Initiating Condition --

SITE AREA EMERGENCY Loss of RPV Inventory Affecting Core Decay Heat Removal Capability.

Operating Mode Applicability:

Cold Shutdown Threshold Values:

(1 OR 2) 1. Loss of Reactor Pressure Vessel (RPV) inventory affecting core decay heat removal capability with CONTAINMENT CLOSURE NOT established as indicated by:

a. RPV level less than 185'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS)

OR b. RPV level CANNOT be monitored for greater than 30 minutes with a possible loss of RPV inventory as indicated by unexplained level rise in ANY of the following:

Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT)

2. Loss of RPV inventory affecting core decay heat removal capability with CONTAINMENT CLOSURE established as indicated by:

a. RPV level less than 181'-10" [TOAF] (63% on Full Range RVLIS)

OR b. RPV level CANNOT be monitored for greater than 30 minutes with a possible loss of RPV inventory as indicated by EITHER of the following:

Unexplained Containment sump level rise.

Erratic Source Range Monitor Indication Basis: CONTAINMENT CLOSURE: Per Operating Procedure 14210-1, Containment Building

Penetrations Verification - Refueling.

Under the conditions specified by this IC, continued lowering in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV.

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 D-51 REV 48 03/08 shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. 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 the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2).

In the cold shutdown mode, normal RCS level and reactor vessel level indication systems (RVLIS) will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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.

If a RVLIS is not available such that the PWR Threshold Value setpoint cannot be determined, then Threshold Value 1.b should be used to determine if the IC has been met.

The 30 minute duration allows sufficient time for actions to be performed to recover needed cooling equipment. For PWRs the effluent release path is not expected with closure established. For BWRs releases would be monitored and escalation would be via Category A ICs if required.

Declaration of a Site Area Emergency is warranted under the conditions specified by the IC.

Escalation to a General Emergency is via CG1 (Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV) or radiological effluent IC RG1 (Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology).

D-52 REV 62 10/14 SYSTEM MALFUNCTION CS2 Initiating Condition --

SITE AREA EMERGENCY Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with

Irradiated Fuel in the RPV.

Operating Mode Applicability:

Refueling Threshold Values:

(1 OR 2) 1. WITH CONTAINMENT CLOSURE NOT established:

a. RPV level less than elevation 185'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS) OR b. RPV level CANNOT be monitored WITH indication of core uncovery as evidenced by ANY of the following:

RE-005 OR 006 40 R/hr RE-0011 3 mR/hr Erratic Source Range Monitor Indication

2. WITH CONTAINMENT CLOSURE established

a. RPV level less than elevation 181'-10" [TOAF] (63% on Full Range RVLIS)

OR

b. RPV level CANNOT be monitored WITH Indication of core uncovery as evidenced by ANY of the following:

RE-005 OR 006 40 R/hr RE-0011 3 mR/hr Erratic Source Range Monitor Indication Basis: CONTAINMENT CLOSURE: Per Operating Procedure 14210-1, Containment Building Penetrations Verification - Refueling.

D-53 REV 48 03/08 Under the conditions specified by this IC, continued lowering in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach or continued boiling in the RPV. 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 or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. 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 the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2).

If a RVLIS is not available such that the PWR Threshold Value setpoint cannot be determined, then Threshold Value 1.b should be used to determine if the IC has been met.

For Threshold Value 2 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 that the ability to monitor level will not be interrupted.

This effluent release is not expected with closure established.

D-54 REV 62 10/14 SYSTEM MALFUNCTION CG1 Initiating Condition --

GENERAL EMERGENCY Loss of RPV Inventory Affecting Fuel Cl ad Integrity with Containment Challenged with Irradiated Fuel in the RPV.

Operating Mode Applicability:

Cold Shutdown Refueling Threshold Values: (1 AND 2 AND 3) 1. Loss of RPV inventory as indicated by unexplained level rise in ANY of the following:

Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT) . AND 2. RPV Level:

a. Less than elevation 181'-10" [TOAF] (63% on Full Range RVLIS) for greater than 30

minutes OR b. RPV level CANNOT be monitored WITH indication of core uncovery for greater than 30 minutes as evidenced by ANY of the following:

RE-005 OR 006 40 R/hr RE-0011 3 mR/hr Erratic Source Range Monitor Indication AND

3. Containment challenged as indicated by ANY of the following:

Explosive mixture inside containment greater than OR equal to 6% H 2 Containment Pressure greater than OR equal to 13 psig WITH CONTAINMENT CLOSURE established greater than OR equal to 52 psig WITH Tech Spec containment integrity intact CONTAINMENT CLOSURE NOT established D-55 REV 48 03/08 Basis: CONTAINMENT CLOSURE: Per Operating Procedure 14210-1, Containment Building Penetrations Verification - Refueling.

For Threshold Value 1 in the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises 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.

For Threshold Value 1 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 that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. For both cold shutdown and refueling modes sump and tank level rises 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.

Threshold Value 2 represents the inability to restore and maintain RPV level to above the top of active fuel. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level.

As water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in up-scaled Containment High Range Monitor indication and possible alarm. Post-TMI studies indicated that 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.

Threshold Values 2a and 2b values are included in the CS2 basis above.

The GE 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 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 GE.

In the context of Threshold Value 3, CONTAINMENT CLOSURE is the action taken to secure containment and its associated structures, syst ems, 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 provi de for re-establishing CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory functions. If the closure is re-established prior to exceeding the temperature or level thresholds of the RCS Barrier and Fuel Clad Barrier Threshold Values, escalation to GE would not occur.

D-56 REV 48 03/08 The site-specific pressure at which CONTAINMENT is considered challenged may change based on the condition of the CONTAINMENT. If the Unit is in the cold shutdown mode and the CONTAINMENT is fully intact then the site-specific setpoint is the CONTAINMENT design pressure (52 psig). This is consistent with typical owner's groups Emergency Response Procedures. With CONTAINMENT CLOSURE established intentionally by the plant staff in preparations for inspection, maintenance, or refueling the setpoint is based on the penetration seals design of 3 psig.

D-57 REV 60 07/13 AC/DC POWERRX and COREALARMS / COMMUNICATIONSNATURAL / DESTRUCTIVEFIRE / EXPLOSIONTOXIC / FLAMMABLESECURITY CR EVACUATIONED DISCRETION SA5-AC power to essential busses reduced to a single power source for greater than 15 minutes.SS2-Failure of RPS Instrumentation to

complete or Initiate an Automatic Reactor Trip Once a RPS Setpoint Has Been Exceeded

ANDManual Trip Was NOT Successful. Modes 1, 2 Only SS4-Complete Loss of Heat Removal CapabilitySU2-Inability to Reach Required Shutdown Within Technical Specification Limits SU8-Inadvertent Criticality Modes 3, 4 Only SU3-UNPLANNED Loss of Most or All Safety System Annunciation or Indication in The Control Room for Greater Than 15 Minutes SU5-RCS Leakage HU1-Natural and Destructive Phenomena

Affecting the PROTECTED AREA HA3-Release of Toxic, Asphyxiant, or Flammable Gases Within or contiguous to a VITAL AREA Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown HA4-HOSTILE ACTION within the

OWNER CONTROLLED AREA or airborne attack threat.

HA5-Control Room Evacuation Has Been

Initiated.SYSTEM MALFUNCTIONS HAZARDSALERT EMERGENCYNOTIFICATION OF UNUSUAL EVENT S I T E A R E A E M E R G E N C Y G E N E R A L E M E R G E N C Y SG1-Prolonged Loss of All Offsite Power ANDProlonged Loss of All Onsite AC Power to essential buses SS1-Loss of All Offsite Power ANDLoss of All Onsite AC Power to Essential Busses SU1-Loss of All Offsite Power to Essential Busses for GREATER THAN 15 Minutes SA2-Failure of RPS Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a RPS System Setpoint Has Been Exceeded AND Manual Trip was Successful. Modes 1, 2, 3

Only SU4-Fuel Clad Degradation SS6-Inability to Monitor a

SIGNIFICANT TRANSIENT in Progress SU6-UNPLANNED Loss of All Onsite

OROffsite Communications Capabilities HU2-FIRE Within PROTECTED AREA Boundary Not Extinguished Within 15 Minutes of Detection HA2-FIRE OR EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown HU3-Release of Toxic, Asphyxiant, OR Flammable Gases Deemed Detrimental to Normal Operation of the Plant HU4 -Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant HG1 -HOSTILE ACTION resulting in loss of physical control of the facilityHS2-Control Room Evacuation Has Been

Initiated ANDPlant Control Cannot Be Established HU5-Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE. HA6-Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert.

HS3-Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency HG2-Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency HA1-Natural and Destructive Phenomena Affecting the Plant VITAL AREA SA4-UNPLANNED Loss of Most or All Annunciation or Indication With EITHER a SIGNIFICANT TRANSIENT in Progress, ORCompensatory Non-Alarming Indicators are Unavailable FISSION PRODUCT BARRIERS FG1-Loss of ANY Two Barriers

ANDLoss or Potential Loss of Third Barrier FS1-Loss or Potential Loss of ANY Two Barriers See Fission Product Barrier Matrix FA1-ANY Loss or ANY Potential Loss of EITHERFuel Clad Barrier ORRCS Barrier See Fission Product Barrier Matrix FU1-ANY Loss ORANY Potential Loss of Containment Barrier See Fission Product Barrier Matrix RAD LEVELS RADIOLOGICALEFFLUENTS RA3Release of Radioactive Material or Rises in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown RA2-Damage to Irradiated Fuel ORLoss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel RA1-Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Effluent Technical Specifications for 15 Minutes or Longer. RU2-Unexpected Rise in Plant Radiation RG1-Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mR TEDE

OR 5000 mR Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology.

RS1-Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR TEDE OR500 mR Thyroid CDE for the Actual or Projected Duration of the Release.

RU1-Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or Longer. HOT INITIATING CONDITIONEMERGENCY ACTION LEVEL MATRIX -MODES 1, 2, 3, and 4 ONLY SG2-Failure of the RPS to complete an Automatic Trip and Manual Trip was NOT successful ANDthere is Indication of an Extreme Challenge to the ability to Cool the Core Modes 1, 2 Only Vogtle Electric Generating PlantSee Fission Product Barrier Matrix SS3-Loss of All Vital DC Power HS4-HOSTILE ACTION within the PROTECTED AREAE-HU1-Damage to a loaded cask CONFINEMENT BOUNDARY D-58 REV 60 07/13 D-59 REV 48 03/08 Fission Product Barri er Degradation NOUE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY Loss or ANY Potential Loss of Containment Op. Modes:

Power Operation, Hot Standby, Startup, Hot Shutdown FA1 ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS Op. Modes:

Power Operation, Hot Standby, Startup, Hot Shutdown FS1 Loss or Potential Loss of ANY Two Barriers Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown FG1 Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown NOTES 1. The logic used for these initiating conditions reflects the following considerations:

• The Fuel Clad Barrier and the RCS Barrier ar e weighted more heavily than the Containment Barrier. NOUE ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

• 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" Threshold Values existed, that, in addition to offsite dose assessments, would require continual assessments of radioac tive inventory and containment integrity. Alternatively, if both Fuel Clad and RCS Barrier "Potential Loss" Threshol d Values 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.

D-60 REV 62 10/14 Vogtle Fission Product Barrier Evaluation General Emergency Site Area Emergency Alert Unusual Event FG1 Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier FS1 Loss or Potential Loss of ANY Two Barriers FA1 ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS FU1 ANY Loss or ANY Potential Loss of Containment Fuel Clad Barrier Loss Potential Loss 1. Critical Safety Function Status Core-Cooling RED

1. Critical Safety Function Status Core Cooling-ORANGE OR Heat Sink-RED

2. Primary Coolant Activity Level Indications of RCS Coolant Activity greater than 300

µCi/gm Dose Equivalent I-131

2. Primary Coolant Activity Level Not Applicable

3. Core Exit Thermocouple Readings Core Exit TCs greater than 1200

°F 3. Core Exit Thermocouple Readings Core Exit TCs greater than 711

°F 4. Reactor Vessel Water Level Not Applicable

4. Reactor Vessel Water Level RVLS LEVEL less than 63%

5. Containment Radiation Monitoring Containment Radiation Monitor RE-005 OR 006 2 6E+5 mR/hr

5. Containment Radiation Monitoring Not Applicable

6. Other Indications Not applicable

6. Other Indications Not applicable 7. Emergency Director Judgment Judgment by the ED that the Fuel Clad Barrier is lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier 7. Emergency Director Judgment Judgment by the ED that the Fuel Clad Barrier is potentially lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier.

RCS Barrier Loss Potential Loss

1. Critical Safety Function Status Not Applicable

1. Critical Safety Function Status RCS Integrity-RED

OR Heat Sink-RED

2. RCS Leak Rate RCS subcooling less than 24

°F {less than 38

° F Adverse} due to an RCS leak greater than Charging / RHR capacity

2. RCS Leak Rate Non-isolable RCS leak (including SG tube Leakage) greater than 120 gpm

3. SG Tube Rupture SGTR resulting in an SI actuation

3. SG Tube Rupture Not Applicable

4. Containment Radiation Monitoring CTMT Rad Monitor RE-005 OR 006 8.7E+2 mR/hr

4. Containment Radiation Monitoring Not Applicable

5. Other Indications Not applicable

5. Other Indications Unexplained level rise in ANY of the following:

Containment sump

Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT) 6. Emergency Director Judgment Judgment by the ED that the RCS Barrier is lost. Consi der conditions not addressed and inability to determine the status of the RCS Barrier 6. Emergency Director Judgment Judgment by the ED that the RCS Barri er is potentially lost. Consider conditions not addressed and inability to determine the status of the RCS Barrier. Containment Barrier Loss Potential Loss

1. Critical Safety Function Status Not Applicable

1. Critical Safety Function Status Containment-RED

2. Containment Pressure Rapid unexplained CTMT pressure lowering following initial pressure rise OR Intersystem LOCA indicated by CTMT pressure or sump level response not consistent with a loss of primary or secondary coolant

2. Containment Pressure CTMT pressure greater than 52 psig OR CTMT hydrogen concentration greater than 6%

OR CTMT pressure greater than 21.5 psig AND Less than the following minimum operable equipment: Four CTMT fan coolers AND One train of CTMT spray

3. Core Exit Thermocouple Reading Not applicable

3. Core Exit Thermocouple Reading CORE COOLING CSF - RED for greater than 15min

OR CORE COOLING CSF - ORANGE for greater than 15min

AND RVLS LEVEL less than 63%

4. SG Secondary Side Release with Primary to Secondary Leakage RUPTURED S/G is also FAULTE D outside of containment OR Primary-to-Secondary leakrate great er than 10 gpm with nonisolable steam release from affected S/G to the environment

4. SG Secondary Side Release with P-to-S Leakage Not applicable 5. CNMT Isolation Valves Status After CNMT Isolation CTMT isolation valve(s)

OR damper(s) are NOT closed resulting in a direct pathway to the environment afte r containment isolation is required

5. CNMT Isolation Valves Status After CNMT Isolation Not Applicable

6. Significant Radioactive Inventory in Containment Not Applicable

6. Significant Radioactive Inventory in Containment CTMT Rad monitor RE-005 OR 006 1.3E+7 mR/hr

7. Other Indications Pathway to the environment exists based on VALID RE-2562C Alarm

AND RE-12444C

OR RE-12442C Alarms

7. Other Indications Not applicable

8. Emergency Director Judgment Judgment by the ED that the CTMT Ba rrier is lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier 8. Emergency Director Judgment Judgment by the ED that the CTMT Ba rrier is potentially lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier D-61 REV 48 03/08 FUEL CLAD BARRIER Threshold Values:

The Fuel Clad Barrier is the zircalloy or stainless steel tubes that contain the fuel pellets.

1. Critical Safety Function Status

______________________________________________________________________

NOTE Heat Sink CSF should not be considered -RED if total AFW flow is less than 570 gpm due to operator action.

______________________________________________________________________

RED path indicates an extreme challenge to the safety function. ORANGE path indicates a

severe challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and

thus these two items indicate potential loss of the Fuel Clad Barrier.

Core Cooling - RED indicates significant superheating and core uncovery and is considered to

indicate loss of the Fuel Clad Barrier.

2. Primary Coolant Activity Level Assessment by the NUMARC EAL Task Force indicate s that this amount of coolant activity is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage.

This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is

considered lost.

There is no equivalent "Potential Loss" Threshold Value for this item.

3. Core Exit Thermocouple Readings Core Exit Thermocouple Readings are included in addition to the Critical Safety Functions to

include conditions when the CSFs may not be in use (initiation after SI is blocked).

The "Loss" Threshold Value of 1200 degrees F corresponds to significant superheating of the coolant. This value corresponds to the temperature reading that indicates core cooling - RED in

Fuel Clad Barrier Threshold Value #1.

The "Potential Loss" Threshold Value of 711 degrees F corresponds to loss of subcooling. This value corresponds to the temperature reading that indicates core cooling - ORANGE in Fuel

Clad Barrier Threshold Value #1.

4. Reactor Vessel Water Level There is no "Loss" Threshold Value corresponding to this item because it is better covered by

the other Fuel Clad Barrier "Loss" Threshold Values.

D-62 REV 62 10/14 The 63% RVLIS value for the "Potential Loss" Threshold Value corresponds to the top of the active fuel. The "Potential Loss" Threshold Value is defined by the Core Cooling - ORANGE

path.

5. Containment Radiation Monitoring The 2.6E+5 mR/hr reading is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment. The reading is 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 into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss Threshold Value #4. Thus, this Threshold Value indicates a loss

of both the fuel clad barrier and a loss of RCS barrier.

There is no "Potential Loss" Threshold Value associated with this item.

7. Emergency Director Judgment This Threshold Value addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier is incorporated in this Threshold Value as a factor in Emergency

Director judgment that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional

information.)

RCS BARRIER Threshold Values:

The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary

isolation valves.

1. Critical Safety Function Status

____________________________________________________________________________

NOTE Heat Sink CSF should not be considered -RED if total AFW flow is less than 570 gpm due to operator action.

____________________________________________________________________________

This Threshold Value uses the Critical Safety Function Status Tree (CSFST) monitoring and functional restoration procedures. An RCS Integrity RED path indicates an extreme challenge to

the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS barrier.

There is no "Loss" Threshold Value associated with this item.

D-63 REV 62 10/14 2. RCS Leak Rate The "Loss" Threshold Value addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in

maintaining RCS pressure and inventory against the mass loss through the leak.

The "Potential Loss" Threshold Value is based on the inability to maintain normal liquid

inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header. A second charging pump being required is indicative of a substantial RCS

leak providing the 120 GPM value.

3. SG Tube Rupture This Threshold Value is intended to address the full spectrum of Steam Generator (SG) tube rupture events in conjunction with Containment Barrier "Loss" Threshold Value #4 and Fuel Clad Barrier Threshold Values. The "Loss" Threshold Value addresses RUPTURED SG(s) for which the leakage is large enough to cause actuation of ECCS (SI). This is consistent to the RCS Barrier "Potential Loss" Threshold Value #2. This condition is described by SGTR resulting in an SI actuation. By itself, this Threshold Value will result in the declaration of an Alert.

However, if the SG is also FAULTED (i.e., two barriers failed), the declaration escalates to a

Site Area Emergency per Containment Barrier "Loss" Threshold Value #4.

There is no "Potential Loss" Threshold Value.

4. Containment Radiation Monitoring The RE-005 OR 006 8.7E+2 mR/hr value indicates the release of reactor coolant to the containment. The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within T/S) into the containment atmosphere. This value is less than that specified for Fuel Clad Barrier Threshold Value #5. Thus, this Threshold Value would be indicative of a RCS leak only. If the radiation monitor reading rise to that specified by Fuel Clad Barrier Threshold Value

1. 5, fuel damage would also be indicated.

There is no "Potential Loss" Threshold Value associated with this item.

5. Other Indications There is no "Loss" Threshold Value associated with this item.

An unexplained level rise in the containment sump, Reactor Coolant Drain Tank or the Waste

Holdup Tank could indicate a RCS leak and is therefore included as a Potential Loss of the

RCS Barrier.

Sump and tank level rises should be evaluated against known or controlled processes which are under way, i.e. draining, filling, venting, etc.

D-64 REV 53 09/09

6. Emergency Director Judgment This Threshold Value addresses any other factors that are to be used by the Emergency

Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this Threshold Value as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power",

for additional information.)

CONTAINMENT BARRIER Threshold Values

The Containment Barrier includes the containment building, its connections up to and including

the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including

the outermost secondary side isolation valve.

1. Critical Safety Function Status This Threshold Value uses Critical Safety Function Status Tree (CSFST) monitoring and

functional restoration procedures. Containment RED path indicates an extreme challenge to the

safety function derived from appropriate instrument readings and/or sampling results, and thus

represents a potential loss of containment. Conditions leading to a containment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this Threshold Value is primarily a discriminator between Site Area Emergency and General Emergency representing a potential

loss of the third barrier.

There is no "Loss" Threshold Value associated with this item.

2. Containment Pressure Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity. Containment pressure and sump levels should rise as a result of the mass and energy release into containment from a LOCA. Thus, sump level or pressure not increasing indicates containment

bypass and a loss of containment integrity.

The 52 PSIG for potential loss of containment is based on the containment design pressure.

Existence of an explosive mixture means a hydrogen and oxygen concentration of at least the lower deflagration limit (greater than 6%) curve exists. The indications of potential loss under this Threshold Value corresponds to some of those leading to the RED path in Threshold Value

1. 1 above and may be declared. As described above, this Threshold Value is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier. The 21.5 PSIG represents a potential loss of containment in that the containment heat removal/depressurization system (e.g., containment sprays, fan coolers, etc.,

but not including containment venting strategies) are either lost or performing in a degraded D-65 REV 48 03/08 manner, as indicated by containment pressure greater than the setpoint at which the equipment was supposed to have actuated.

3. Core Exit Thermocouples In this Threshold Value, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing. For units using the CSF status trees a direct correlation to those status trees can be made if the effectiveness of the restoration procedures is also evaluated as stated below.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be ineffective. The reactor vessel level chosen should be consistent with the emergency response guides applicable to the facility.

The conditions in this potential loss Threshold Value represent an imminent core melt sequence which, if not corrected, could lead to vessel failure and an raised potential for containment failure. In conjunction with the Core Cooling and Heat Sink criteria in the Fuel and RCS barrier columns, this Threshold Value would result in the declaration of a General Emergency -- loss of two barriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path.

There is no "Loss" Threshold Value associated with this item.

4. SG Secondary Side Release With Primary To Secondary Leakage This "loss" Threshold Value recognizes that SG tube leakage can represent a bypass of the containment barrier as well as a loss of the RCS barrier. The "loss" Threshold Value addresses the condition in which a RUPTURED steam generator is also FAULTED. This condition represents a bypass of the RCS and containment barriers. In conjunction with RCS Barrier "loss" Threshold Value #3, this would always result in the declaration of a Site Area Emergency.

The other leakage "loss" Threshold Value addresses SG tube leaks that exceed 10 gpm in conjunction with a nonisolable release path to the environment from the affected steam generator. The threshold for establishing the nonisolable secondary side release is intended to be a prolonged release of radioactivity from the RUPTURED steam generator directly to the environment. This could be expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e., SGTR with concurrent loss of offsite power and the RUPTURED steam generator is required for plant cooldown or a stuck open relief valve). If the main condenser is available, there may be rel eases via air ejectors, gland seal exhausters, and other similar controlled, and often monitored, pathways. These pathways do not meet the intent of a nonisolable release path to the environment. These minor releases are assessed using Abnormal Rad Levels / Radiological Effluent ICs.

D-66 REV 62 10/14 Users should realize that the two "loss" Thresholds described above could be considered redundant. This was recognized during the development process. The inclusion of a Threshold that uses Emergency Procedure commonly used terms like "ruptured and faulted" adds to the

ease of the classification process and has been included based on this human factor concern.

5. Containment Isolation Valve Status After Containment Isolation This Threshold Value addresses incomplete containment isolation that allows direct release to

the environment. It represents a loss of the containment barrier.

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 noble gases. Typical filters have an efficiency of 95-99% 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.

There is no "Potential Loss" Threshold Value associated with this item.

6. Significant Radioactive Inventory in Containment The 1.3E+7 mR/hr value indicates significant fuel damage well in excess of the Threshold Values associated with both loss of Fuel Clad and loss of RCS Barriers. As stated in Section 3.8, a major release of radioactivity requiring offsite 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 less than 20%. A

radiation monitor reading corresponding to 20% fuel clad damage is specified here.

There is no "Loss" Threshold Value associated with this item.

7. Other Indications This Threshold Value covers other leakage pathways that indicate loss of the containment barrier. Leakage from the Containment would be routed through various ventilation systems where the specific monitors would indicate a release. Increased activity on RE-2562C in conjunction with an alarm on RE-12444C OR RE-12442C would indicate a breach in containment. An alarm on RE-2562-C alone would indicate increased activity in containment but not necessarily a breach in containment. An alarm on RE-12444C OR RE-12442C alone would indicate activity being released to the environment but not necessarily from containment. This threshold demonstrates a breach in containment by establishing radioactivity in containment and a clear release path to the environment.

D-67 REV 62 10/14 8. Emergency Director Judgment This Threshold Value addresses any other factors that are to be used by the Emergency

Director in determining whether the Containment barrier is lost or potentially lost. In addition, the

inability to monitor the barrier should also be incorporated in this Threshold Value as a

factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.

(See also IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC

Power", for additional information.)

D-68 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Natural and Destructive Phenomena Affecting the PROTECTED AREA.

Operating Mode Applicability:

All Threshold Value: (1 OR 2 OR 3 OR 4 OR 5 OR 7) 1. Confirmed Seismic monitoring alarm indicates seismic event OR plant operators report an earthquake was "felt".

2. Report by plant personnel of tornado OR high winds greater than 100 mph striking within PROTECTED AREA.

3. Crash of vehicle, large enough to cause significant damage, into plant structures containing functions or systems required for safe shutdown within the PROTECTED AREA boundary.

4. Report by plant personnel of an unanticipated EXPLOSION within the PROTECTED AREA boundary resulting in VISIBLE DAMAGE to permanent structure OR equipment.

5. Report of turbine failure resulting in casing penetration OR damage to turbine OR generator seals.

6. Not applicable.

7. Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours in accordance with 11889-C, Severe Weather Checklist.

Basis: PROTECTED AREA: the area which normally encompasses all controlled areas within the security protected area fence.

EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts energy of sufficient fo rce to potentially damage permanent structures, systems, or components.

VISIBLE DAMAGE: is 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 safe ty structure, system , or component. Example 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.

D-69 REV 48 03/08 These ICs are categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. Areas identified in the Threshold Values define the location of the event based on the potential for damage to equipment contained therein. Escalation of the event to an Alert occurs when the magnitude of the event is sufficient to result in damage to equipment contained in the specified location.

Threshold Value #1 - 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. For most plants with seismic instrumentation, the seismic switches are set at an acceleration of about 0.01g.

Threshold Value #2 is based on the assumption that a tornado striking (touching down) or high winds within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. The high wind 100 mph value is based on FSAR design basis (110 mph design) as the highest meter reading available. If such damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert. Threshold Value #3 addresses crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant VITAL AREA, the event may be escalated to Alert. For Threshold Value #4 only those EXPLOSIONs of sufficient force to damage permanent structures or equipment within the PROTECTED AREA should be considered. No attempt is made in this Threshold Value to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The Emergency director also needs to consider any security aspects of the EXPLOSION, if applicable.

Threshold Value #5 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. 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 and HU3. Generator seal damage observed after generator purge does not meet the intent of this Threshold Value because it did not impact normal operation of the plant. This Threshold Value is consistent with the definition of a NOUE while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure or in conjunction with a steam generator tube rupture. These latter events would be classified by the radiological ICs or Fission Product Barrier ICs.

Threshold Value #6 addresses the effect of flooding caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. PROBABILISTIC D-70 REV 48 03/08 Threshold Value #7 covers site-specific phenomena of the hurricane based on the severe weather mitigation procedure.

D-71 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT FIRE Within PROTECTED AREA Boundary No t Extinguished Within 15 Minutes of Detection.

Operating Mode Applicability:

All Threshold Value:

1. FIRE in buildings OR areas contiguous to any of the following areas NOT extinguished within 15 minutes of control room notification OR verification of a control room alarm unless disproved by personnel observation within 15 minutes of the alarm:

Containment Building NSCW Cooling Towers Diesel Generator Building Auxiliary Building Fuel Handling Building Control Building Diesel Fuel Oil Storage Tank Pumphouse Auxiliary Feedwater Pumphouse Basis: FIRE: is combustion characterized by heat and li ght. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIREs. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA: the area which normally encompasses all controlled areas within the security protected area fence.

The purpose of this IC is to address the magnitude and extent of FIREs that may be potentially significant precursors to damage to safety systems. 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 VALID fire detection system alarm. Verification of a fire detection system alarm includes actions that can be taken with the control room or other nearby site-specific location to ensure that the alarm is not spurious. A verified 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.

D-72 REV 56 09/10 The time frames for evaluation and classification are concurrent beginning with the credible notification that a fire is occurring or a VALID fire detection system 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). The site-specific list should be limited and applies to buildings and areas contiguous (in actual contact with or immediately

adjacent) to plant VITAL AREAs or other significant buildings or areas. The intent of this IC is not to include buildings (i.e., warehouses) or areas that are not contiguous (in actual contact

with or immediately adjacent) to plant VITAL AREAs. This excludes FIREs within administration buildings, waste-basket FIREs, and other small FIREs of no safety consequence.

D-73 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Release of Toxic, Asphyxiant, or Flammable Gases Deemed Detrimental to Normal Operation of the Plant.

Operating Mode Applicability:

All Threshold Values:

(1 OR 2) 1. Report OR detection of toxic, asphyxiant, OR flammable gas that has OR could enter the Owner Controlled Area in amounts greater than life threatening OR flammable concentrations that can affect NORMAL PLANT OPERATIONS.

2. Report by Local, County, OR State Officials for evacuation OR sheltering of site personnel based on an offsite toxic, asphyxiant, OR flammable gas event.

Basis: 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.

This IC is based on the existence of uncontrolled releases of toxic, asphyxiant, or flammable gas that may enter the Owner Controlled Area and affect normal plant operations. It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. Offsite events are included through a warning by local officials as the resultant affect on NORMAL PLANT OPERATIONS would be the same. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation. The Threshold Values are not intended to require significant assessment or quantification. The IC assumes an uncontrolled process that has the potential to affect plant operations, or personnel safety.

D-74 REV 55 03/10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed Security CONDITION or threat which indicates a potential degradation in the level of safety of the plant.

Operating Mode Applicability:

All Threshold Values: (1 OR 2 OR 3) 1. A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as reported by security shift supervision.

2. A CREDIBLE VEGP security THREAT notification.

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

Basis: Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs.

CREDIBLE THREAT: A threat is considered credible through use of 90321-C, Threat

Assessment and Security Force Protection Recommendations SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and /or intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under HA4, HS4 and HG1.

D-75 REV 55 03/10 A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. The licensee shall consider upgrading the emergency response status and emergency classification level in accordance with the site's Safeguards Contingency Plan and Emergency Plan.

In Threshold 1, reference is made to security shift supervision 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 Safeguards Contingency Plan.

This threshold is based on site specific securi ty plans. Site specific Safeguards Contingency Plans are based on guidance provided by NEI 03-12.

The intent of Threshold Value 2 is to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a CREDIBLE THREAT.

The intent of Threshold Value 3 is to ensure that notifications for the security threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. It is not the intent of this EAL to replace existing non-hostile related EALs involving aircraft.

This Threshold Value is met when a plant receives information regarding an aircraft threat from the 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 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 would be via HA4 would be appropriate if the threat involves an airliner within 30 minutes of the plant.

D-76 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU5 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Other Conditions Existing Which in t he Judgment of the Emergency Director Warrant Declaration of a NOUE.

Operating Mode Applicability:

All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process OR have occurred which indicate a potential degradation of the level of safety of the plant OR indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response OR monitoring are expected unless further degradation of safety systems occurs.

Basis: This Threshold Value is intended to address 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 NOUE emergency class.

From a broad perspective, one area that may warrant Emergency Director judgment is related to likely or actual breakdown of site-specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel.

D-77 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA1 Initiating Condition -- ALERT Natural and Destructive Phenomena A ffecting the Plant VITAL AREA.

Operating Mode Applicability:

All Threshold Values: (1 OR 2 OR 3 OR 4 OR 5 OR 6) 1. Seismic monitoring system confirms seismic event greater than OBE Earthquake acceleration of 0.12 g.

2. Tornado OR high winds greater than 100 mph within the PROTECTED AREA boundary AND resulting in VISIBLE DAMAGE to any of the following plant structures/equipment OR the Control Room has indication of degraded performance of those systems:

Containment Building Fuel Handling Building NSCW Cooling Towers Control Building Diesel Generator Building Diesel Fuel Oil Storage Tank Pumphouse Auxiliary Building Auxiliary Feedwater Pumphouse

3. Vehicle crash within PROTECTED AREA boundary AND resulting in VISIBLE DAMAGE to any of the following plant structures OR equipment therein OR control indication of degraded performance of those systems:

Containment Building Fuel Handling Building NSCW Cooling Towers Control Building Diesel Generator Building Diesel Fuel Oil Storage Tank Pumphouse Auxiliary Building Auxiliary Feedwater Pumphouse

4. Turbine failure-generated missiles result in any VISIBLE DAMAGE to OR penetration of any area containing safety-related equipment, their controls OR their power supplies.

Containment Building Fuel Handling Building NSCW Cooling Towers Control Building Diesel Generator Building Diesel Fuel Oil Storage Tank Pumphouse Auxiliary Building Auxiliary Feedwater Pumphouse

5. Uncontrolled flooding in areas of the plant that creates an industrial safety hazard (e.g., electric shock) that precludes access nece ssary to operate or monitor safety equipment.

6. Sustained hurricane winds greater than 74 mph onsite resulting in VISIBLE DAMAGE to plant structures within the PROTECTED AREA boundary containing equipment necessary for safe shutdown, or has caused damage as evidenced by control room indication of degraded performance of those systems.

D-78 REV 48 03/08 Basis: PROTECTED AREA: the area which normally encompasses all controlled areas within the security protected area fence.

VISIBLE DAMAGE: is 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 safe ty structure, system , or component. Example 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.

The Threshold Values in this IC escalate from the NOUE Threshold Values in HU1 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 cont rol 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 Threshold Value to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other ICs (e.g., System Malfunction).

Threshold Value #1 is based on the OBE earthquake FSAR design basis of 0.12g horizontal acceleration. Seismic events of this magnitude can result in a plant VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. See EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, for information on seismic event categories.

Threshold Value #2 is based on the FSAR design basis. Wind loads greater than 110 mph can cause damage to safety functions. 100 mph was used for the Threshold Value because it is the highest reading available in the Control Room.

Threshold Value #3 addresses crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant. Threshold Value #4 addresses the threat to safety related equipment imposed by missiles generated by main turbine rotating component failures. This list of areas includes all areas containing safety-related equipment, their controls, and their power supplies. This Threshold Value is, therefore, consistent with the definition of an ALERT in that if missiles have damaged or penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the plant.

Threshold Value #5 addresses the effect of internal flooding that has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to operate or monitor safety equipment represents a potential for substantial degradation of the level of safety of the plant. This flooding may have been caused D-79 REV 48 03/08 by internal events such as component failure s, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant, that are not designed to be wetted or submerged.

Threshold Value #6 covers site-specific phenomena of a hurricane. The Threshold Value is based on damage attributable to the wind.

D-80 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2 Initiating Condition -- ALERT FIRE OR EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.

Operating Mode Applicability:

All Threshold Value:

1. FIRE OR EXPLOSION AND affected system parameter indications show degraded performance OR plant personnel report VISIBLE DAMAGE to permanent structures OR safety related equipment in any of the following VITAL AREAs: . Containment Building Fuel Handling Building NSCW Cooling Towers Control Building Diesel Generator Building Diesel Fuel Oil Storage Tank Pumphouse Auxiliary Building Auxiliary Feedwater Pumphouse Basis: FIRE: is combustion characterized by heat and li ght. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIREs. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts energy of sufficient fo rce to potentially damage permanent structures, systems, or components.

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.

VISIBLE DAMAGE: is 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 safe ty structure, system , or component. Example 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.

The areas listed contain functions and systems required for the safe shutdown of the plant to determine if the FIRE or EXPLOSION is potentially affecting any redundant trains of safety systems. Escalation to a higher emergency class, if appropriate, will be based on System D-81 REV 48 03/08 Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radiological Effluent, or Emergency Director Judgment ICs.

This Threshold Value addresses a FIRE / EXPLOSION and not the degradation in performance of affected systems. System degradation is addressed in the System Malfunction Threshold Values. The reference to damage of systems is used to identify the magnitude of the FIRE / EXPLOSION and to discriminate against minor FIREs / EXPLOSIONs. The reference to safety systems is included to discriminate against FIREs / 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 that the FIRE / EXPLOSION was large enough to cause damage to these systems. Thus, the designation of a single train was intentional and is appropriate when the FIRE / EXPLOSION is large enough to affect more than one component.

This situation is not the same as removing equipment for maintenance that is covered by a plant's Technical Specifications. Removal of equipment for maintenance is a planned activity controlled in accordance with procedures and, as such, does not constitute a substantial degradation in the level of safety of the plant. A FIRE / EXPLOSION is an UNPLANNED activity and, as such, does constitute a substantial degradation in the level of safety of the plant. In this situation, an Alert classification is warranted.

The inclusion of a "report of VISIBLE DAMAGE" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this Threshold Value to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform these damage assessments. The Emergency Director also needs to consider any security aspects of the EXPLOSIONs, if applicable.

D-82 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA3 Initiating Condition -- ALERT Release of Toxic, Asphyxiant or Flamm able Gases Within or Contiguous to a VITAL AREA Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

Operating Mode Applicability:

All Threshold Values: (1 OR 2) 1. Report OR detection of toxic OR asphyxiant gas within OR contiguous to a VITAL AREA in concentrations that may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).

2. Report OR detection of flammable gases in concentration greater than the LOWER FLAMMABILITY LIMIT within OR contiguous to a VITAL AREA.

Basis: 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.

IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH): A condition that either poses an immediate threat to life and health or an immediate threat of severe exposure to contaminants which are likely to have adverse delayed effects on health.

LOWER FLAMMABILITY LIMIT (LFL): The minimum concentration of a combustible substance that is capable of propagating a flame through a homogenous mixture of the combustible and a gaseous oxidizer.

This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and areas contiguous to plant VITAL AREAs or other significant buildings or areas (i.e., service water pump house). The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not contiguous or immediately adjacent to plant VITAL AREAs. It is appropriate that augmented monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radioactive Effluent, or Emergency Director Judgment ICs.

Threshold Value #1 is met if measurement of toxic gas concentration results in an atmosphere that is IDLH within a VITAL AREA or any area or building contiguous to VITAL AREA.

D-83 REV 48 03/08 Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas.

Threshold Value #2 is met when the flammable gas concentration in a VITAL AREA or any building or area contiguous to a VITAL AREA exceed the LOWER FLAMMABILITY LIMIT. 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 Threshold Value addresses concentrations at which gases can ignite/support combustion. 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. Once it has been determined that an uncontrolled release is occurring, then sampling must be done to determine if the concentration of the released gas is within this range.

D-84 REV 55 03/10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4 Initiating Condition -- ALERT HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat. Operating Mode Applicability:

All Threshold Values:

1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLED AREA as reported by security shift supervision.

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

Basis: Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs.

HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and /or intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

OWNER CONTROLED AREA - The utility owned property around the plant where access is controlled during declared emergencies by the plant security force.

These EALs 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).

D-85 REV 55 03/10 Threshold 1 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, hunt ers, or physical disputes between employees within the OCA. Those events are adequately addressed by other EALs.

Note that this EAL is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes ISFSI's that may be outside the PROTECTED AREA but still within the OWNER CONTROLLED AREA.

Threshold 2 addresses the immediacy of an expected threat arrival or impact on the site within a relatively short time.

The intent of this EAL is to ensure that notifications for the airliner attack threat are made in a timely manner and that OROs 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.

This EAL is met when a plant receives information regarding an airliner attack threat from NRC and the airliner is within 30 minutes of the plant. Only the plant to which the specific threat is made need declare the Alert.

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.

D-86 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 Initiating Condition -- ALERT Control Room Evacuation Has Been Initiated.

Operating Mode Applicability:

All Threshold Value:

1. Entry into 18038 - Operation From the Remote Shutdown Panels, procedure for Control Room evacuation.

Basis: With the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facility is necessary. Inability to establish plant control from outside the control room will escalate this event to a Site Area Emergency.

D-87 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6 Initiating Condition -- ALERT Other Conditions Existing Which in t he Judgment of the Emergency Director Warrant Declaration of an Alert.

Operating Mode Applicability:

All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process OR have occurred which involve actual OR likely 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 intentional malicious dedicated efforts of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Basis: HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and /or intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

This Threshold Value is intended to address 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.

D-88 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition - SITE AREA EMERGENCY Control Room Evacuation Has Been Initiated AND Plant Control Cannot Be Established.

Operating Mode Applicability:

All Threshold Value:

1. a. Control Room evacuation has been initiated AND b. Control of the plant CANNOT be established per 18038 - Operation From the Remote Shutdown Panels, within 15 minutes.

Basis: Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. The intent of this IC is to capture those events where control of the plant cannot be reestablished in a timely manner. Site-specific time for transfer based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes without additional justification. The determination of whether or not control is established at the remote shutdown panel is based on Emergency Director (ED) judgment. The ED is expected to make a reasonable, informed judgment within the site-specific time for transfer that the licensee has control of the plant from the remote shutdown panel.

The intent of the Threshold Value 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. These safety functions are r eactivity control, RCS inventory, and secondary heat removal.

D-89 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS3 Initiating Condition - SITE AREA EMERGENCY Other Conditions Existing Which in t he Judgment of the Emergency Director Warrant Declaration of Site Area Emergency.

Operating Mode Applicability:

All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process 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 that exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Basis: HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and /or intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

This Threshold Value is intended to address 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.

D-90 REV 55 03/10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS4 Initiating Condition - SITE AREA EMERGENCY HOSTILE ACTION within the PROTECTED AREA Operating Mode Applicability:

All Threshold Value: 1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by security shift supervision.

Basis: HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and /or intimidate the licensee to achieve an end. This

includes attack by air, land, or water using weapons , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be

included. HOSTILE ACTION should not be construed to include acts of civil disobedience or

felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs

should be used to address such activities, (e.g., violent acts between individuals in the owner

controlled area.)

PROTECTED AREA: the area which encompasses all controlled areas within the security protected area fence.

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 prem ised 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 ORO 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, hunt ers, or physical disputes between employees within the PROTECTED AREA. Those events are adequately addressed by other EALs.

D-91 REV 55 03/10 Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack.

D-92 REV 55 03/10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 Initiating Condition - GENERAL EMERGENCY HOSTILE ACTION resulting in loss of physical control of the facility.

Operating Mode Applicability:

All Threshold Value: (1 OR 2) 1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions.

2. 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.

Basis: HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and /or intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons , explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

IMMINENT: Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur.

This EAL 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, RCS inventory, and secondary heat removal. If control of the plant equipment necessary to maintain safety functions can be transferred to

another location, then the above initiating condition is not met.

Threshold Value 2 should also address loss of physical control of spent fuel pool cooling systems if imminent fuel damage is likely for a freshly off-loaded reactor core in pool). If the calculated SFP "time to boil" is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or less, spent fuel damage is likely.

D-93 REV 48 03/08 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG2 Initiating Condition - GENERAL EMERGENCY Other Conditions Existing Which in t he Judgment of the Emergency Director Warrant Declaration of General Emergency.

Operating Mode Applicability:

All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that 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 Ac tion Guideline exposure levels offsite for more than the immediate site area.

Basis: HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, takes 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 the NPP. Non- terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

This Threshold Value is intended to address 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.

D-93a REV 60 07/13 HAZARDS AND OTHE R CONDITIONS AFFECTING PLANT SAFETY E-HU1 Initiating Condition Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: Not Applicable Threshold Value: (1) 1. Damage to a loaded dry fuel storage cask CONFINEMENT BOUNDARY due to natural phenomena events, accident conditions OR any condition in the opinion of the emergency director that affects OR causes a loss of loaded dry fuel storage cask CONFINEMENT BOUNDARY.

Basis: CONFINEMENT BOUNDARY: is the barrier(s) between areas containing radioactive substances and the environment.

A NOUE in this IC 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.

Any condition, which, in the judgment of the Emergency Director, is a potential degradation in the level of safety of the ISFSI. Emergency director judgment is to be based on known conditions and the expected response to mitigati ng activities within a short time period.

D-94 REV 48 03/08 SYSTEM MALFUNCTION SU1 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

1. a. Loss of power to OR from transformers 1(2)NXRA AND 1(2)NXRB resulting in loss of all off-site electrical power to BOTH 1(2)AA02 AND 1(2)BA03 for greater than 15 minutes AND b. Emergency diesel generators supplying power to BOTH 1(2)AA02 AND 1(2)BA03.

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. Plants that have the capability to cross-tie AC power from a companion unit may take credit for the redundant power source in the associated Threshold Value for this IC. Inability to effect the cross-tie within 15 minutes warrants declaring a NOUE.

D-95 REV 48 03/08 SYSTEM MALFUNCTION SU2 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT Inability to Reach Required Shutdown Within Technical Specification Limits.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

1. Plant is NOT brought to required operating mode within Technical Specifications LCO Action Statement Time limit.

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 one hour report under 10 CFR 50.72 (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 NOUE 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 a NOUE is based on the time at which the LCO-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 System Malfunction, Hazards, or Fission Product Barrier Degradation ICs.

D-96 REV 48 03/08 SYSTEM MALFUNCTION SU3 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

1. UNPLANNED loss of most OR all of the MCB annunciators OR indicators associated with safety systems for greater than 15 minutes.

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC and its associated Threshold Value are 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 (e.g., SPDS, plant computer, etc.).

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a greater 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 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 Threshold Value 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 via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE is based on SU2 "Inability to Reach Required Shutdown Within Technical Specification Limits." The annunciators or indicators for this Threshold Value include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other Threshold Values (e.g., area, process, and/or effluent rad monitors, etc.).

D-97 REV 48 03/08 Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.

This NOUE will be escalated to an Alert if a transient is in progress during the loss of annunciation or indication.

D-98 REV 62 10/14 SYSTEM MALFUNCTION SU4 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT Fuel Clad Degradation.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown

Threshold Values:

(1 OR 2) 1. RE-48000 radiation monitor reading greater than 5.0 µCi/gm indicating fuel clad degradation greater than Technical Specification allowable limits.

2. RCS coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits as indicated by ANY of the following:

Dose Equivalent I-131 greater than 1 Ci/gm for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Dose Equivalent I-131 greater than Technical Specification figure 3.4.16-1 limits RCS specific activity greater than 100/E Ci/gm gross radioactivity Basis: VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC is included as a NOUE because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. EALThreshold Value #1 addresses the site specific radiation monitor reading that provides indication of loss of fuel clad

integrity.

The Threshold Value #2 addresses coolant samples exceeding coolant technical specifications for iodine spike. Escalation of this IC to the Alert level is via the Fission Product Barrier Degradation Monitoring ICs. Though the referenced Technical Specification limits are mode dependent, it is appropriate that the Threshold Value's be applicable in all modes, as they indicate a potential degradation in the level of safety of the plant. The companion IC to SU4 for the Cold

Shutdown/Refueling modes is CU5.

D-99 REV 48 03/08 SYSTEM MALFUNCTION SU5 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT RCS Leakage.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Values:

(1 OR 2) 1. Unidentified OR pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Basis: This IC is included as a NOUE 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 and 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). The Threshold Value 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 IC to the Alert level is via Fission Product Barrier Degradation ICs.

D-100 REV 48 03/08 SYSTEM MALFUNCTION SU6 Initiating Condition --

NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of All Onsite OR Offsite Communications Capabilities.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Values:

(1 OR 2) 1. UNPLANNED loss of ALL of the following on-site communications capability affecting the ability to perform routine operations:

In plant telephones Public address system Plant radio systems

2. UNPLANNED loss of ALL of the following off-site communications capability:

ENN (Emergency Notification Network)

ENS (Emergency Notification System) Commercial phones (Radio, PBX, Satellite, Wireless)

VOIP (Voice Over Internet Protocol) OPX (Off Premise Extension)

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

The purpose of this IC and its associated Threshold Values 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 offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

D-101 REV 48 03/08 The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant conditions. This Threshold Value is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being used to make communications possible.

The list for onsite communications loss encompasses the loss of all means of routine communications. The list for offsite communications loss encompasses the loss of all means of communications with offsite authorities.

D-102 REV 48 03/08 SYSTEM MALFUNCTION SU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent Criticality.

OPERATING MODE APPLICABILITY Hot Standby Hot Shutdown Threshold Value:

1. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC addresses inadvertent criticality events. While t he primary concern of this IC is 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), the IC is applicable in other modes in which inadvertent criticalities are possible. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated). The Cold Shutdown/Refueling IC is CU8.

This condition is identified using the startup rate monitor. The term "sustained" is used in order to allow exclusion of expected short term positive startup rates from planned control rod movements. These short term positive startup rates are the result of the rise in neutron population due to subcritical multiplication.

D-103 REV 48 03/08 SYSTEM MALFUNCTION SA2 Initiating Condition --

ALERT Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protecti on System Setpoint Has Been Exceeded AND Manual Trip Was Successful.

Operating Mode Applicability:

Power Operation Startup Hot Standby Threshold Value:

______________________________________________________________________

NOTE A successful manual trip for purposes of declaration is any action taken from the Main Control Board (MCB) that rapidly inserts the control rods, this can be accomplished by tripping the reactor using the Reactor Trip switches on the MCB OR by de-energizing both Rod Drive Motor Generator sets from the MCB. ______________________________________________________________________

1. Indication(s) exist that a reactor protection setpoint was exceeded and an automatic scram did not occur, and a manual scram resulted in the reactor being subcritical.

Basis: This condition indicates failure of the automatic protection system to trip 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 and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified here because failure of the automatic protection system is the issue. A manual reactor trip is considered to be a trip input to the automatic Reactor Protection System. Failure of manual scram would escalate the event to a Site Area Emergency.

The Reactor should be considered subcritical when reactor power level has been reduced to less than 5% power and SUR is negative.

D-104 REV 48 03/08 SYSTEM MALFUNCTION SA4 Initiating Condition --

ALERT UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With EITHER (1) a SIGNIFICANT TRANSIENT in Progress, OR (2) Compensatory Non-Alarming Indicators are Unavailable.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

(1 AND EITHER 2.a OR b.) 1. UNPLANNED loss of most OR all MCB annunciators OR indicators associated with safety systems for greater than 15 minutes AND EITHER 2. a. A SIGNIFICANT TRANSIENT is in progress OR b. Compensatory non-alarming indications are NOT available Basis: UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following: (1) automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Trip, (4) Safety Injection Activation, or (5) thermal power oscillations >10%.

This IC and its associated Threshold Values are 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 transient. Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a greater 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 D-105 REV 48 03/08 also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide augmented 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 Threshold Value 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 via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE is based on SU2 "Inability to Reach Required Shutdown Within Technical Specification Limits." The annunciators or indicators for this Threshold Value include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other Threshold Values (e.g., area, process, and/or effluent rad monitors, etc.).

"Compensatory non-alarming indications" in this cont ext includes computer based information such as SPDS. This should include all computer systems av ailable for this use depending on specific plant design and subsequent retrofits. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required.

Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes, no IC is indicated during these modes of operation.

D-106 REV 48 03/08 SYSTEM MALFUNCTION SA5 Initiating Condition --

ALERT AC power capability to essential busses reduc ed to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout.

Operating Mode Applicability:

Power Operation Startup Hot Shutdown Threshold Value:

1. a. AC power capability to 1(2)AA02 AND 1(2)BA03 reduced to a single power source for greater than 15 minutes AND b. ANY additional single failure will result in station blackout.

Basis: This IC and the associated Threshold Values are intended to provide an escalation from IC SU1, "Loss of All Offsite Power To Essential Busses for Greater Than 15 Minutes." The condition indicated by this IC is the degradation of the offsite and onsite power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of offsite power with a concurrent failure of one emergency generator to supply power to its emergency busses. Another related condition could be the loss of all offsite power and loss of onsite emergency diesels with only one train of emergency busses being backfed from the SAT, or the loss of onsite emergency diesels with only one train of emergency busses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with IC SS1, "Loss of All Offsite and Loss of All Onsite AC Power to Essential Busses." The Threshold Values allow credit for operation of installed design features, such as cross-ties or swing diesels, provided that abnormal or emergency operating procedures address their use. Consider the impact of this condition on other shared safety functions.

D-107 REV 48 03/08 SYSTEM MALFUNCTION SS1 Initiating Condition --

SITE AREA EMERGENCY Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

1. Loss of all AC power indicated by:

a. Loss of power to OR from transformers 1(2)NXRA AND 1(2)NXRB resulting in loss of all off-site electrical power to BOTH 1(2)AA02 AND 1(2)BA03 for greater than 15 minutes AND b. Failure of emergency diesel generators to supply power to emergency busses.

AND c. Restoration of at least one 4160V ESF bus, AA02 OR BA03, has NOT occurred within 15 minutes of time of loss of all AC power Basis: Loss of all AC power compromises all plant safe ty systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. The 15 minute time duration is selected to exclude transient or momentary power losses.

Escalation to General Emergency is via Fission Product Barrier Degradation or IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power." Consideration should be given to operable loads nece ssary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable. If this bus was the only energized bus then a Site Area Emergency per SS1 should be declared.

D-108 REV 48 03/08 SYSTEM MALFUNCTION SS2 Initiating Condition --

SITE AREA EMERGENCY Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protecti on System Setpoint Has Been Exceeded AND Manual Trip Was NOT Successful.

Operating Mode Applicability:

Power Operation Startup Threshold Value:

______________________________________________________________________

NOTE A successful manual trip for purposes of declaration is any action taken from the Main Control Board (MCB) that rapidly inserts the control rods, this can be accomplished by tripping the reactor using the Reactor Trip switches on the MCB OR by de-energizing both Rod Drive Motor Generator sets from the MCB. ______________________________________________________________________

1. Indications exist that a reactor protection system setpoint was exceeded and automatic trip did not occur, and a manual trip did not result in the reactor being made subcritical.(NOTE)

Basis: Automatic and manual trip are not considered successful if action away from the reactor control console was required to trip the reactor.

The Reactor should be considered subcritical when reactor power level has been reduced to less than 5% power and SUR is negative.

Under these continued power generation conditions, the reactor may be producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that may lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Barrier Degradation or Emergency Director Judgment ICs.

D-109 REV 48 03/08 SYSTEM MALFUNCTION SS3 Initiating Condition --

SITE AREA EMERGENCY Loss of All Vital DC Power.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

1. Loss of Vital DC power to 125 VDC Buses AD1, BD1, CD1, AND DD1 indicated by bus voltage indications less than 105 VDC for greater than 15 minutes.

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.

105 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads.

D-110 REV 48 03/08 SYSTEM MALFUNCTION SS4 Initiating Condition --

SITE AREA EMERGENCY Complete Loss of Heat Removal Capability.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

_____________________________________________________________________

NOTE Heat Sink CSF should not be considered -RED if total AFW flow is less than 570 gpm due to operator action. _____________________________________________________________________

1. Complete Loss of Heat Removal Capability as indicated by:

a. Core Cooling CSF - ORANGE AND b. Heat Sink CSF - RED Basis: This Threshold Value addresses complete loss of functions, including ultimate heat sink (NSCW), required for hot shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other Threshold Values.

Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency would be via Abnormal Rad Levels / Radiological Effluent, Emergency Director Judgment, or Fission Product Barrier Degradation ICs.

D-111 REV 48 03/08 SYSTEM MALFUNCTION SS6 Initiating Condition --

SITE AREA EMERGENCY Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

1. a. SIGNIFICANT TRANSIENT in progress AND b. Loss of most OR all of the MCB annunciators OR indicators associated with safety systems AND c. Compensatory non-alarming indications are NOT available AND d. Indications needed to monitor the Critical Safety Function Status Tree parameters are NOT available Basis: SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following: (1) automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Trip, (4) Safety Injection Activation, or (5) thermal power oscillations >10%.

This IC and its associated Threshold Value are intended to recognize the inability of the control room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public.

The annunciators for this Threshold Value are limited to include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other Threshold Values (e.g., rad monitors, etc.)

"Compensatory non-alarming indications" in this cont ext includes computer based information such as SPDS. The indications needed to monitor safety functions necessary for protection of the public must include control room indications, computer generated indications and dedicated annunciation capability. The specific indications are those used to determine such functions as the ability to shut down the reactor, D-112 REV 48 03/08 maintain the core cooled, to maintain the reactor coolant system intact, and to maintain containment intact. "Planned" 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 of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a greater 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 Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide augmented monitoring of system operation.

D-113 REV 48 03/08 SYSTEM MALFUNCTION SG1 Initiating Condition --

GENERAL EMERGENCY Prolonged Loss of All Offsite Power AND Prolonged Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability:

Power Operation Startup Hot Standby Hot Shutdown Threshold Value:

(1 AND EITHER 2 OR 3) 1. Loss of ALL AC power indicated by:

a. Loss of power to OR from transformers 1(2)NXRA AND 1(2)NXRB resulting in loss of all off-site electrical power to BOTH 1(2)AA02 AND 1(2)BA03 for greater than 15 minutes AND b. Failure of emergency diesel generators to supply power to emergency busses.

AND EITHER

2. Restoration of at least one 4160V ESF bus, AA02 OR BA03, within 4 hr. of time of loss is NOT likely OR 3. Fuel Clad Barrier Evaluation indicates continuing degradation (Loss or Potential Loss) due to core cooling Basis: Loss of all AC power compromises all plant safe ty systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore AC power is based on a site blackout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, "Station Blackout". Appropriate allowance for offsite emergency response including evacuation of surrounding areas should be considered. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

D-114 REV 48 03/08 This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency 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.

D-115 REV 48 03/08 SYSTEM MALFUNCTION SG2 Initiating Condition --

GENERAL EMERGENCY Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip was NOT Successful AND there is Indication of an Ex treme Challenge to the Ability to Cool the Core.

Operating Mode Applicability:

Power Operation Startup Threshold Value:

______________________________________________________________________

NOTE Heat Sink CSF should not be considered - RED if total AFW flow is less than 570 gpm due to operator action. ______________________________________________________________________

1. Indications exist that a reactor protection system setpoint was exceeded and automatic trip did not occur, and a manual trip did not result in the reactor being made subcritical.

AND Core Cooling CSF - RED OR Heat Sink CSF - RED Basis: Automatic and manual trip are not considered successful if action away from the reactor control console is required to trip the reactor.

The Reactor should be considered subcritical when reactor power level has been reduced to less than 5% power and SUR is negative.

Under the conditions of this IC and its associated Threshold Values, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as emergency boration, the continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence. This Threshold Value equates to a Subcriticality RED condition.

The extreme challenge to the ability to cool the core is intended to mean that the core exit temperatures are at or approaching 1200 degrees F or that the reactor vessel water level is below the top of active fuel. This Threshold Value equates to a Core Cooling RED condition.

D-116 REV 48 03/08 Another consideration is the inability to initially remove heat during the early stages of this sequence. If emergency feedwater flow is insufficient to remove the amount of heat required by design from at least one steam generator, an extreme challenge should be considered to exist. This Threshold Value equates to a Heat Sink RED condition.

In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (typ ically 3 to 5% power) 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 matrix declaration to permit maximum offsite intervention time.

E-1 REV 61 06/14 E. NOTIFICATION METHODS AND PROCEDURES This section describes the plan for notification of onsite and offsite Vogtle Electric Generating Plant (VEGP) response

personnel and State, local, and Nuclear Regulatory Commission (NRC) emergency response centers. Actual methods and

sequencing of notifications are covered in appropriate

implementation procedures (NMP-EP-111, Emergency

Notifications, and Procedure 91204-C, Emergency Response

Communications). Tables E-1 and E-2 and figure E-1 present

the initial notification concept.

E.1 NOTIFICATION OF PERSONNEL The emergency director is responsible for classifying an event (section D) into the appropriate emergency class and

then notifying onsite and offsite personnel accordingly.

This notification will involve sounding the appropriate plant

emergency alarm signal, making appropriate announcements over

the plant public address system, activation of the recall

system and using appropriate plant telephone system.

The primary means for notification of personnel within the

protected area is the public address (PA) system. Upon

declaration of a Notification of Unusual Event (NUE), an

Alert, a Site Area Emergency, or a General Emergency, the

emergency director will order an announcement of the

emergency

The tone signals for each of these classes of emergency

conditions, as well as for a fire, are as follows:

• NUE: announcement only, no tone signal.

• Alert: warble tone.

• Site Area Emergency: warble tone.

• General Emergency: warble tone.

• Fire: siren tone.

During security related events the ED may elect to not sound a warning tone and, in such cases, will provide event specific instructions for onsite personnel over the PA system as well as other available communications means as needed.

The supervisor nuclear security is responsible for notifying

Plant Wilson, the training center, the visitors center, and

recreation park staff. All visitors at the visitors center

will leave the site if directed by the emergency director or

if a Site Area or General Emergency is declared. Security

will activate the site siren to notify personnel on site, outside the protected area of an evacuation order.

E-2 REV 60 07/13 The Security Department will also be responsible for evacuating all visitors and nonessential personnel from the

Plant Vogtle Recreational Park and for the verification of

the evacuation of all nonessential personnel from Plant

Wilson, the training center, and the remaining areas inside

the owner controlled area.

Visitors within the protected area are escorted by a

permanently badged individual. This individual is

responsible for informing the visitors of emergencies when

they occur and for taking action to evacuate the visitors

from the site, as necessary.

Plant and contractor personnel will be trained on actions to

be taken in an emergency prior to their work assignment.

Otherwise, they must be escorted by an individual who has

been trained in emergency procedures. The training includes

instructions on the methods of personnel notification and the

required personnel actions in the event of an emergency.

Notification of the corporate staff is performed in

accordance with NMP-EP-002 / NMP-GM-036.

The notification procedure includes notification of Emergency

Response Organization Personnel (ERO) not on site. ERO

members will be notified by means of an autodialer system

activated by on-shift personnel. In addition to those

personnel recalled; Operations, Maintenance, and Security

personnel required to report shall be contacted by on-shift

personnel from their own respective department.

E.2 NOTIFICATION OF STATE AND LOCAL RESPONSE PERSONNEL The emergency director is responsible for the completion of the Initial Message Form (see figure E-1 for a sample form)

and for the notification of the following within 15 min of

the declaration of an emergency:

• Georgia Emergency Management Agency Emergency

Operations Center (EOC) communicator.

• Burke County Emergency Operations Center (EOC)

communicator.

• South Carolina warning point.

• Aiken County sheriff dispatcher.

• Barnwell County sheriff dispatcher.

• Allendale County central dispatch.

E-3 REV 56 09/10

• Department of Energy-Savannah River (DOE SR) Operations Center communicator.

These agencies will be responsible for notifying appropriate

response personnel in accordance with their emergency plans and

procedures. A dedicated telephone system, known as the Emergency Notification Network (ENN), will normally be used to accomplish these notifications. Section F describes the ENN

and backup means of communication. Figure E-1 presents the

sample initial message form for making notifications to these

response centers. This form has been developed in conjunction

with appropriate offsite agencies. The message form may be transmitted electronically or verbally.

E.3 NOTIFICATION OF FEDERAL AGENCIES The emergency director is responsible for ordering notification calls to the DOE-SR Operations Center by ENN and to the NRC

Operations Center by the Emergency Notification System (ENS) or

commercial telephone as backup within prescribed time

constraints from the declaration of an emergency. Examples of

the type of Initial Emergency Message Form used to provide the

initial notification to the SRS Operations Center, and the NRC

Operations Center Event Notification Form used for the NRC

notification, are shown in figures E-1 and E-2, respectively.

E.4 NOTIFICATION OF THE PUBLIC It is the responsibility of VEGP to provide adequate means for notifying the public, or to be assured that such means are

provided. In case of an emergency, State and local agencies

are responsible for activating the alert notification system.

Administrative and physical means have been established for

providing early initial warning and subsequent clear

instructions to the populace within the plume exposure pathway

emergency planning zone (EPZ). The alert notification system, except for the Savannah River Site (SRS), is described in

appendix 3; this system has the capability to complete the

initial alert notification of residents within the plume EPZ in

about 15 min. Follow up messages can be delivered to the public

over commercial broadcast

In the event that an emergency is declared at the Vogtle

Electric Generating Plant (VEGP), DOE-SR has agreed to provide

for the prompt notification of all persons on the SRS within

VEGP's plume exposure pathway EPZ. See appendix 5.

VEGP will provide offsite authorities with supporting

information for their messages to the public. Such messages, consistent with the emergency classification scheme, will

provide the public with instructions in regard to specific

protective actions to be taken by occupants of affected areas.

E-4 REV 56 09/10 E.5 FOLLOWUP EMERGENCY MESSAGES The emergency director is responsible for the completion of a followup emergency message. (See figure E-1, sample form.)

The appropriate support coordinator will ensure the emergency

communicator(s) periodically provide followup messages to the

appropriate offsite Federal, State, and local authorities.

E.6 VERIFICATION OF NOTIFICATION MESSAGES All notification messages must be verified. When the ENN is used, verification is accomplished by roll call. This is a

suitable mechanism, since the ENN is a multiparty dedicated

telephone line.

When commercial telephone or radio is used for notification, the called party will contact the Vogtle Electric Generating

Plant to verify the validity of the message or use the

authentication system provided by the State of South Carolina

EPD. When electronic notifications are performed, verification is accomplished by roll call.

REV 35 6/02 TABLE E-1 (SHEET 1 OF 2)

INITIAL NOTIFICATION SYSTEM NORMAL WORKING HOURS Responsible Communicator Primary Notification System Party Notified Control Room Staff PA system 1. Protected area and Admin.

Bldg. personnel PA system telephone, pager 2. Vogtle duty manager Autodialer system and

beeper system 3. Recall for emergency response personnel not on

site Security Department Plant 1. Visitors center personnel telephone 2. Training center personnel system 3. Personnel onsite, outside protected area

4. Personnel at Plant Wilson Direct contact Recreation area occupants ENN Communicator ENN 1. GEMA Operations Center communicator

2. Burke County EOC communicator

3. SRS Operations Center communicator

4. South Carolina Warning Point 5. Aiken County sheriff dispatcher

6. Barnwell County sheriff dispatcher

7. Allendale County Central Dispatcher REV 58 09/12 TABLE E-1 (SHEET 2 OF 2)

Responsible Communicator Primary Notification System Party Notified Operations Staff ENS NRC Operations Center Vogtle duty manager telephone Nuclear Duty Officer

REV 58 09/12 TABLE E-2 (SHEET 1 OF 2)

INITIAL NOTIFICATION SYSTEM BACKSHIFT HOURS Responsible Communicator Primary Notification System Party Notified Control Room Staff Public 1. Protected area address personnel system Telephone 2. Vogtle duty manager and pager Autodialer system and

beeper system 3. Recall for emergency response personnel not on

site ENN Communicator ENN 1. GEMA Operations Center communicator

2. Burke County EOC communicator 3. SRS communicator 4. South Carolina Warning Point 5. Aiken County sheriff dispatcher 6. Barnwell County sheriff dispatcher 7. Allendale County Central dispatcher Operations Staff ENS 1. NRC Operations Center Vogtle duty manager Telephone 1. Nuclear Duty Officer VEGP security Plant 1. Visitors center personnel telephone system 2. Training center personnel 3. Personnel onsite, outside protected area 4. Plant Wilson

REV 33 12/01 TABLE E-2 (SHEET 2 OF 2)

Responsible Communicator Primary Notification System Party Notified VEGP Security Direct 1. Recreation area occupants (cont) contact FIGURE E-1 EXAMPLE OF INITIAL EMERGENCY MESSAGE AFOR STATE AND LOCAL RESPONSE AGENCIES REV 43 05/06

F-1 REV 56 09/10 F. EMERGENCY COMMUNICATIONS This section describes the provisions for communication among the principal response organizations and among the Vogtle

Electric Generating Plant (VEGP) Emergency Response

Facilities (ERF). See Procedure 91204-C, Emergency Response

Communications, for details of communication systems and

operation. The communications described in this section are

summarized in table F-1.

F.1 COMMUNICATIONS WITH THE STATE OF GEORGIA AND BURKE COUNTY F.1.1 State of Georgia

The primary means of communication between the VEGP and the State of Georgia is the Emergency Notification

Network (ENN), a dedicated telephone system from the

plant to the State Emergency Operating Center (EOC) at

Georgia Emergency Management Agency (GEMA)

headquarters in Atlanta, Georgia. There is also an

ENN terminal at GEMA's FEOC. Extensions for this

system are located in the control room, technical

support center (TSC), emergency operations facility (EOF). The ENN system is available and manned 24 h

per day. Commercial telephones and a Southern Company

Communications in Atlanta provide backup for the

dedicated telephone circuits. The plant telephone

backup power is supplied by a battery system.

An Administrative Decision Line (ADL) connects the EOF, SRS Operations Center, the GEMA FEOC, the SEOCs

of both states and the three South Carolina counties.

This is a prearranged conference call over commercial

telephone lines and may be used for discussions other

than emergency notifications.

F.1.2 Burke County

The primary means of communication between VEGP and Burke County is the ENN, a dedicated telephone system

from the plant to the Burke County EOC. Commercial

telephones and Southern Company Communications provide backups for the ENN.

The ENN is available and manned 24 h per day. At the plant, the emergency director is in charge of

communications to the Burke County EOC. Actual

communications may be completed by others as

designated by the emergency director. At the Burke

County EOC, the Burke County emergency management

director is in charge of communications.

F-2 REV 43 05/06 F.2 COMMUNICATIONS WITH THE STATE OF SOUTH CAROLINA AND AIKEN, BARNWELL, AND ALLENDALE COUNTIES F.2.1 State of South Carolina

The primary means of communication between the VEGP and South Carolina is the ENN, a dedicated telephone

system from the plant to South Carolina emergency

response agencies. The ENN has multiple drops in

South Carolina state facilities which are located at:

the South Carolina Warning Point (the State emergency

operations center (SEOC)) which is manned 24 h per

day. Commercial telephones provide the backup for the

ENN.

An ADL connects the EOF, the SRS Operations Center, the GEMA FEOC, the SEOCs of both states, and the three

South Carolina counties. This prearranged conference

call over commercial lines may be used for discussions

other than emergency notifications.

The State emergency preparedness director is responsible for communication at the SEOC with the

plant, the Savannah River Site, and contiguous local

and State governments.

F.2.2 Aiken, Barnwell, and Allendale Counties

The primary means of communication between VEGP and the South Carolina counties is the ENN, a dedicated

telephone system which includes the plant and Aiken, Barnwell, and Allendale County emergency response

agencies. Commercial telephone is the backup means of

communication.

Each county has 2 ENN drops: one at the County EOC and one at the following 24-h warning points:

County Location Aiken Sheriff Department (Dispatcher) Allendale Allendale County Communications Barnwell 911 Center)

F.3 COMMUNICATIONS WITH THE SAVANNAH RIVER SITE The primary means of communication between VEGP and the Savannah River Site (SRS) is the ENN. SRS has two ENN drops, both located in their Operations Center.

The ENN system is available and manned 24 h per day.

Commercial telephones provide a backup for the ENN.

F-3 REV 30 07/00 At VEGP, the emergency director is in charge of communications to the SRS Operations Centers. Actual communications may be completed by others as designated

by the emergency director. At the SRS Operations

Center, the DOE duty officer is in charge of

communications with VEGP.

F.4 COMMUNICATIONS WITH NUCLEAR REGULATORY COMMISSION AND OTHER FEDERAL AGENCIES The primary means of communication between VEGP and the Nuclear Regulatory Commission (NRC) is the emergency notification system (ENS). The ENS phone service is

provided by the Federal Telecommunications System (FTS).

Emergency notification system phones are located in the

control room, TSC, and EOF. The NRC also provides the

health physics network (HPN). The HPN phone service is

also provided by the FTS. Health physics network phones

are located in the TSC and EOF. In the TSC, the HPN

telephone is located in the communications room and will

be attended by VEGP personnel until an NRC

representative arrives. The NRC Region II office in

Atlanta, Georgia, may also be connected on the ENS

through Rockville, Maryland.

The Emergency Response Data System (ERDS) is the primary means by which the transmission of plant parameters

occurs. The ERDS computer, when activated, periodically

transmits a predefined list of critical plant parameters

over the dedicated ERDS FTS lines to the NRC Operations

Center in Rockville, Maryland.

Commercial telephone lines and the Southern Company Communications serve as backups to the ENS and HPN.

Communications with other Federal emergency response

organizations will be by telephone.

F.5 COMMUNICATIONS AMONG VEGP EMERGENCY RESPONSE FACILITIES Communications among the control room, TSC, OSC, and EOF can be completed using dedicated telephone circuits, normal plant telephones, and radio, using the plant

network. The radio system is also used for

communications with the radiological monitoring teams.

Communications available at each emergency response

facility are as follows:

1. CONTROL ROOM

• Dedicated telephone circuits to the TSC, EOF, and

OSC (one for each location).

• The ENN.

• The NRC ENS.

F-4 REV 28 03/99

• Normal plant phones.

• In-plant radio console

• Sound-powered phones.

• Plant page system.

• Commercial dial.

• Southern Company Communications.

• Facsimile.

2. TECHNICAL SUPPORT CENTER

• Dedicated telephone circuits to the control room, EOF, and OSC (one to each location).

• The ENN.

• The ENS.

• Two FTS ERDS lines.

• The HPN.

• Facsimile.

• Normal plant phones.

• In-plant radio.

• Sound-powered phones.

• Plant page system.

• Field team radio remote.

• Additional FTS lines.

F-5 REV 57 06/12

• Commercial dial.

• South Carolina radio remote.

• Southern Company Communications.

3. OPERATIONS SUPPORT CENTER

• Dedicated voice telephone circuits to the EOF and TSC (one for each location).

• Normal plant phones.

• In-plant radio transceiver.

• Plant page system.

• Commercial dial.

4. EMERGENCY OPERATIONS FACILITY

• Dedicated telephone circuits to the control room, OSC, and TSC (one each to the control room, OSC, TSC).

• The ENN.

• The ENS.

• The HPN.

• Facsimile.

• Normal plant phones.

• Commercial dial.

• Additional FTS lines.

F-6 REV 60 07/13

• Southern Company Communications.

• JIC hotline.

• Field team radio remote.

• ADL prearranged conference.

5. JOINT INFORMATION CENTER

• Southern Company Communications.

• GPC general office dial.

• Commercial dial.

• VEGP dial.

• Facsimile.

F.6 MEDICAL SUPPORT FACILITY COMMUNICATIONS Communications with Doctors Hospital or the Burke Medical Center will be by commercial telephone. Radio contact through the Burke County EOC serves as a backup. The Burke

County Ambulance Service is equipped with a radio for

communications with the hospitals. The ambulance service and

hospitals within the state are interconnected in a statewide

hospital radio network. Vogtle Electric Generating Plant can

communicate with the ambulances by contacting the hospitals, which have radio communications with the ambulances.

F-7 REV 43 05/06 F.7 ALERTING EMERGENCY RESPONSE PERSONNEL As described in section E, notification of onsite personnel at VEGP will be completed through a combination of public address system announcements, tone signals, and

proceduralized telephone calls. After normal working hours, VEGP personnel not on site at the time of the emergency will

be notified by beeper (for plant management) or by telephone

call using an autodialer system.

F.8 COMMUNICATIONS SYSTEM TESTS Communication channels with the State of Georgia, Burke County, the State of South Carolina, Aiken County, Barnwell County, Allendale County, SRS, and the NRC will be tested

monthly, using the extensions in the control room, TSC, and EOF. Communications systems that link the control room, TSC, EOF, State EOC's and GEMA FEOC, County EOC's and SRS EOC are

tested quarterly. The communication system for communicating

between the TSC, EOF, and VEGP field monitoring teams is

tested quarterly. Communications procedures and systems are

also tested biennially during a communications drill. This

drill is normally conducted during the biennial exercise.

Emergency Response Data System computers are tested once per quarter, or as dictated by NUREG 1394.

F.9 VEGP RADIOLOGICAL MONITORING TEAMS In-plant monitoring teams will communicate with the Health Physics or OSC communicator at least every half hour per Procedure 91302-C, In-Plant Sampling and Surveys. Field

monitoring teams will communicate with the EOF or TSC

communicator at least every half hour per Procedure 91303-C, Field Sampling and Surveys.

There are multiple radio frequencies used for communicating with monitoring teams. Transmitters and antennas are located

throughout the owner controlled area for field monitoring

teams and the in-plant monitoring teams. The field

monitoring team radio covers the entire plume exposure

pathway EPZ. Remote stations for communicating with the

field monitoring teams are located in the TSC and the EOF.

F-8 REV 60 07/13 TABLE F-1 EMERGENCY RESPONSE COMMUNICATIONS

SUMMARY

CONTROL ROOM TSC OSC EOF JOINT INFORMATION CENTER CORPORATE OFFICE BURKE COUNTY EOC GEMA FEOC GEMA EOC SRS NRC OPERATIONS CENTER NRC REGION II OFFICE

ATLANTA SC STATE EOC SC COUNTY EOC'S SC COUNTY WARNING POINTS ENN ENS (a) VEGP DIAL DEDICATED DIAL BELL DIAL IN-PLANT RADIO PLANT PAGE SYSTEM (PA)

SOUTHERN COMPANY COMMUNICATIONS HPN FACSIMILE FIELD TEAM RADIO DNR RADIO SC EMD RADIO ADMINISTRATIVE DECISION LINE ERDS (a) ENS available when bridged with NRC Operations Center G-1 REV 24 03/97 G. PUBLIC EDUCATION AND INFORMATION The detailed planning for public information actions during an emergency, including rumor control, is contained in the

Emergency Communications Plan (appendix 8). A general

description of the public education and information program

follows.

Georgia Power Company (GPC) and Southern Nuclear Operating Company (SNC) in coordination with State and local officials will provide information to the public at least annually

regarding how they will be notified and what their actions

should be in an emergency. All materials used to provide

emergency planning information to the public (information

brochures, advertisements, signs and notices, etc.) will be

reviewed by GPC and SNC on an annual basis. All materials will be updated as necessary and printed material distributed

annually.

The means for disseminating this information include

information on siren poles, signs, notices in public areas, and publications distributed at least annually.

Information is distributed annually to residents in the plume

exposure pathway emergency planning zone (EPZ) through an

emergency information calendar entitled, "The Plant Vogtle

Emergency Information Calendar". The text of the calendar

addresses the following subjects:

Summary - What to do if you are warned of an

emergency at Plant Vogtle. Emergency response plans. When an emergency will affect you. How you will be told about an emergency. What actions you might need to take. What if you are told to shelter. What if you are told to evacuate. Steps to be prepared for an emergency. Steps for using map and chart of evacuation area. Steps to take to get ready for the trip. Classes of accidents.

G-2 REV 41 12/05 Where to get more information or other help. What radiation is. Special needs card. A joint message from Southern Nuclear Operating Company, the states of Georgia and South Carolina, and Burke, Aiken, Allendale, and Barnwell counties

explaining the contents of the calendar, asking that

they read the emergency information, requesting that

they keep it in a handy place, and giving contacts

for further information.

In addition, public information personnel provide public education programs to the community. These programs typically focus on plant operational concepts, plant safety

considerations, and radiation.

G.1 INFORMATION FOR TRANSIENTS Signs and notices providing information to transients are placed in public recreation areas as well as other public places in the plume EPZ such as siren poles, the

Plant Vogtle Visitors Center, and in commercial

establishments; for example, motels, restaurants, and

gas stations. This material will include the following

information:

How people will be warned of an emergency. What to do if warned of an emergency. A list of radio and television stations to tune to

for further information.

G-3 REV 60 07/13 Finally, a Vogtle emergency information brochure will be made available within the EPZ to transients at commercial establishments, churches, motels, hunting

clubs, Creek and Cawden Plantations, the Plant Vogtle

Visitors Center, and through residents whose land is

used by nonresidents (e.g., the occasional nonresident

hunter). Outside the EPZ, the brochure will be made

available to timber company offices for distribution to

their employees who enter the EPZ on Company business, and to the Waynesboro Office of the Agriculture

Stabilization & Conservation Service for distribution to

farmers who farm, but do not reside, in the EPZ.

G.2 JOINT INFORMATION CENTER OPERATIONS The joint information center (JIC) will be the principal point of contact with the news media during an emergency. The JIC will accommodate public information

representatives from SNC, GPC, State, local and Federal

response agencies. News releases and media briefings

will be coordinated to the maximum extent possible.

GPC will utilize the Corporate Headquarters Building at 241 Ralph McGill Boulevard, NE, Atlanta, Georgia, to

serve as a temporary information center until the JIC in

Waynesboro is activated. Once activated, the JIC becomes the principal location for dissemination of

information about the emergency. This facility is

located approximately 15 miles from the plant and can

accommodate a large number of reporters.

Principal GPC and SNC contacts for the media will be the public information director and the designated Company

spokesperson. The Company spokesperson position is

filled by individuals who, under normal operations, hold

supervisory positions on the SNC Corporate or plant

staff and are technically and professionally qualified

to perform this important function. The Company

spokesperson has access to all information and telephone

contact with the emergency director through the EOF

Manager. He briefs the media on plant status and

Company emergency activities. In addition, technical

briefers have been designated who can provide general

and background information, as appropriate, to reporters

at the JIC.

Further information on the public information organization and information flow to the public during

an emergency is available in appendix 8.

G-4 REV 60 07/13 G.3 OFFSITE AGENCY COORDINATION GPC and SNC will provide timely and accurate information to local, State and federal agencies and will seek reciprocal information from these agencies. Efforts

will be made to coordinate periodic press briefings and

to issue public statements in conjunction with these

Government agencies. A joint public information center

operation at the Joint Information Center will provide ample opportunity for all parties represented to review

all information prior to public release.

G.4 RUMOR CONTROL Providing timely, accurate, and consistent information to the public is considered the best way to dispel rumors. Rumors are controlled by having a single source

of information. In an emergency, a rumor control

network will be activated. News media will be monitored

to detect and respond to misinformation. The public

will be instructed to listen to their radio or TV.

Offsite information is the responsibility of offsite

agencies; however, rumor control will be coordinated

between the states, SNC, and GPC. The states, SNC, and

GPC provide information jointly to the rumor control

desk at the JIC.

G.5 MEDIA EDUCATION GPC will offer an annual program to acquaint the news media with the methodology for obtaining information about overall emergency preparedness at Vogtle.

Training will include information about the plant, radiation, and the role of the emergency news center.

H-1 REV 51 10/08 H. EMERGENCY FACILITIES AND EQUIPMENT Following the declaration of an emergency, response activities will be coordinated at a number of facilities.

These Emergency Response Facilities (ERF) and the equipment

which will be used for accident assessment and monitoring

functions are described in this section.

H.1 EMERGENCY FACILITIES

1. TECHNICAL SUPPORT CENTER (TSC) The TSC has been established consistent with NUREG 0696, as described below. The TSC is shared by both units and is located adjacent to the Unit 1 control

room at el 220 ft. The layout of the TSC and

location of key personnel are shown in figure H-1.

Walking time from the TSC to the control room is

less than 2 min. The TSC covers about 5900 ft 2 and can accommodate 25 people. It is an integral part

of the Seismic Category I control building

structure.

The TSC provides plant management and technical support personnel (including five Nuclear Regulatory

Commission (NRC) personnel) with a facility from

which they can assist plant operating personnel

located in the control room during an emergency.

The emergency director and NRC director will be

located in close proximity to ensure proper communications. The TSC is equipped with a computer

system, which provides source term and

meteorological data and technical data displays to

allow TSC personnel to perform detailed analysis and

diagnosis of abnormal plant conditions, including

assessment of any significant release of

radioactivity to the environment. In addition, the

TSC has ready access to plant records, some of which

are stored in the TSC and some in the control room. The TSC structure and ventilation system are designed to ensure that the TSC personnel are

protected from radiological hazards. The TSC

ventilation is a separate system and not a part of

the control room ventilation system. The air supply is filtered by high efficiency particulate air (HEPA) and charcoal filters. During

accident conditions the filtration system provides

filtered cooling air meeting the following exposure

design criteria: 5 rem - whole body; 30 rem -

thyroid; and 75 rem - skin dose for a 30 day

occupancy.

H-2 REV 23 4/96 The air filtration system will be placed in service when required by Procedure 91110-C, Duties of the Health Physics Supervisor (TSC). The ventilation system can be manually controlled from the TSC. The air filtration system will also be placed in the

filtration mode automatically whenever the control

room HVAC is isolated. The air filtration system, when activated, automatically processes all of the

outdoor air together with approximately 25 percent of

the recirculated air through the filtration unit. In

the initial actuation of the air filtration system, a

parallel signal is initiated to deenergize the onsite

technical support center battery room and toilet

exhaust fans and dampers. The continuous influx of

outdoor air with no positive exhaust is designed to

pressurize the area to 0.125 in. WG. After the

filtration process, the filtered air is mixed with the

balance of recirculated air for further conditioning

by the normal supply system. When the TSC is activated, there will be a portable radiation monitor placed in the TSC to alert personnel

of the presence of high radiation levels. In

addition, portable radiation monitors are available

for personnel in transit from the TSC to other areas.

Portable air breathing apparatus and anti-

contamination clothing are provided in the TSC. The TSC normal lighting is supplied from normal offsite power through a motor control center backed up

by the security diesel generator. The TSC essential

lighting is supplied from the essential lighting

system. Self-contained, battery-operated emergency

lighting is provided as a backup to the normal

lighting for ingress and egress only and is located in

the TSC hallway. Power for TSC vital equipment is provided from either the motor control center backed up by the security

diesel generator or from a battery-backed

uninterruptible power supply system. Power to the dc

system is provided via battery chargers, one of which

is powered from this same motor control center. The records storage is shared by the control room and the TSC. Separate copies of the following documents

will be stored in the control room complex and in the

TSC:

• Technical Specifications.

• Plant Operating Procedures.

• Final Safety Analysis Report.

H-3 REV 57 06/12

• Emergency Plan.

• Emergency Plan Implementing Procedures.

The following documents are available in Vogtle Electric Generating Plant (VEGP) Document Control:

• Plant operating records.

• Plant Review Board records and reports.

• System piping and instrumentation diagrams and heating, ventilation, and air-conditioning (HVAC)

flow diagrams.

• Piping area drawings.

• Electrical one-line, elementary, and wiring

diagrams.

• Control logic and loop diagrams.

• Technical Specifications.

• Final Safety Analysis Report.

• Emergency Plan.

• Records needed to perform the functions of the emergency operating facility (EOF) when it is not operational. The above records are available in current form and are updated as necessary to ensure currency and

completeness. Operations at this facility are directed by the TSC manager. 2. OPERATIONS SUPPORT CENTER (OSC)

The OSC has been established to be consistent with NUREG 0696 as described below. The OSC is located on the second floor of the maintenance building in the

lunch room. The lunch room is located in the

southeast corner of the second floor and is accessed

via the south stairwell or the east stairwell. Figure

H-2 defines the OSC layout and shows access

stairwells. The OSC is where operational support

personnel (such as instrument technicians, engineers, mechanics, electricians, chemical/radiation

technicians, equipment operators, incoming shift

personnel, etc.) assemble to aid in the response to an

emergency.

H-4 REV 30 07/00 Preparations for dispatching maintenance teams to the plant will include the following steps as described in Procedure 91104, Duties of the OSC Manager, and

Procedure 91202, Activation and Operation of the OSC.

A. Members of the team will be selected by the OSC manager based on the type of work to be conducted.

B. The work will be preplanned by the maintenance personnel.

C. Drawings, manuals, and other procedures will be obtained from the document control room or

maintenance building. If the equipment manuals are

not available in either facility, a complete set of

manuals is available in the service building.

D. Dosimetry will be obtained from the OSC emergency kit or at the HP control point.

E. Protective clothing and equipment, radios, and other supplies will be obtained from the emergency kit

stored in the OSC or the health physics control

point. All equipment will be checked before leaving

the OSC or health physics control point.

F. A radiation work permit (RWP) will be completed in accordance with standard practices at the health

physics control point, TSC, or OSC. Standard

procedures include emergency RWPs. Procedure

91301-C, Emergency Exposure Guidelines, provides

instruction for emergency exposures.

G. The tools and equipment needed to conduct repairs and take corrective actions will be determined. This

equipment may be available in the auxiliary building

tool crib or in the maintenance building.

Arrangements for tools and equipment not located

onsite will be made in coordination with the support

coordinator.

H. ALARA and job briefings will be held with each team in the OSC, the TSC, or at the health physics

control point, as appropriate. Work to be

performed, special precautions, plant conditions, and radiological information will be included in the

briefings.

I. Upon completion of the job, the team members will be debriefed and their exposures recorded.

H-5 REV 61 06/14 Status boards containing plant conditions and emergency classification will be available in the OSC. Emergency kits containing radiation monitoring equipment, first aid supplies, decontamination supplies, breathing apparatus, portable lighting, and

hand-held radios are stored in the OSC. Emergency kit

contents are listed in appendix 4. In the event that this facility becomes uninhabitable, the functions of the OSC will be conducted from

Clearance and Tagging (C&T) located in the control

building. Evacuation of the OSC will be conducted

according to Procedure 91202-C, Activation and

Operation of the OSC. This Procedure describes the

method by which the OSC is evacuated and the movement

of personnel to other facilities in an orderly manner.

The OSC manager will keep the TSC manager informed of

the initiation, progress, and completion of the

evacuation and relocation of the OSC personnel. Operations at this facility are directed by the OSC manager.

3. EMERGENCY OPERATIONS FACILITY The Emergency Operations Facility is described in Appendix 7.

4. ALTERNATIVE FACILITY During a security related event or other event that precludes onsite access, the TSC and OSC ERO staff will be directed to an alternative facility. This facility is located in the Plant Vogtle JIC building.

The alternative facility is located in the rear conference room of the JIC and is equipped with the necessary communications and data links to support communications with the control room, site security, and the EOF. The available communications and data links also provide access to SNC document management resources, work planning resources for performing engineering assessment activities, including damage control team planning and preparation for return to the site. Procedural guidance for this facility is provided in NMP-EP-135, Alternative Facility Setup and Operation.

H-6 REV 38 05/05

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H-7 REV 38 05/05

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H-8 REV 57 06/12 H.2 NEWS CENTER FACILITIES The Joint Information Center is described in appendix 8.

H.3 ACTIVATION AND STAFFING OF EMERGENCY FACILITIES During the initial stages of an emergency situation, emergency activities at VEGP are directed from the control room. For a Notification of Unusual Event, no

other facilities need be activated.

For security related events, the activation of emergency facilities may be delayed as described in section B.

H-9 REV 38 05/05 Upon declaration of an Alert or higher level classification, the TSC will be activated and will be operational within about an hour of the initial

notification. Overall direction and control will be

exercised from the TSC for an Alert or higher declaration.

Activation of the OSC will be initiated at an Alert or higher level classification. Support personnel will be

directed to report to that facility as appropriate for

the specific situation. The OSC will be operational

within about an hour of initial notification.

Activation of the EOF is described in Appendix 7.

H.4 PLANT MONITORING AND DATA HANDLING SYSTEMS

1. GEOPHYSICAL PHENOMENA MONITORS

a. Meteorological A meteorological monitoring program is in place at VEGP. Instruments are mounted on a 60-m tower

located to the south-southwest of the power

block. Parameters measured and transmitted to the

control room include:

• Windspeed (10 m and 60 m).

• Wind direction (10 m and 60 m).

• Standard deviation of horizontal wind direction

(10 m).

• Vertical temperature difference (10 m and

60 m).

• Ambient temperature (10 m).

• Dewpoint temperature (10 m).

• Precipitation (base).

H-10 REV 35 6/02 An equipment building which houses the recording, calibration, and amplification equipment is located near the base of the tower. The system is powered

by an uninterruptible power supply consisting of wet

cell batteries, charger, and inverter for high

availability. The important parameters for characterizing the transport of airborne radioactivity are windspeed, wind direction, and atmospheric stability (derived

from the standard deviation of the horizontal wind

direction or vertical temperature difference). These

meteorological parameters are used in a

calculational methodology to assess the offsite

radiological consequences of accidental releases of

airborne radioactivity. The methodology is

described in section I, Accident Assessment.

b. Hydrologic The normal source of plant cooling water is the Savannah River, which provides makeup to the cooling

towers. The probable maximum flood level has been

determined to be about 140 ft mean sea level (MSL).

However, since the access elevations to safety-

related structures are at 220 ft MSL, high river

level is not relevant to plant safety. The ultimate

heat sink for VEGP is the nuclear service cooling

water towers. Two 100-percent towers are provided

for each unit, and the system will provide

sufficient shutdown cooling for approximately 30

days with no makeup. Because of these design

features, hydrologic monitors will not be required

for initiation of emergency actions; therefore, there will be no emergency levels based on

hydrologic monitors.

c. Seismic Seismic monitoring instrumentation for VEGP consists of time-history accelerographs, central control unit, and free-field ETNA. A strong-motion accelerometer (SMA) is installed in the containment tendon gallery on the basemat. The

second SMA is located on the containment operating floor at elevation 220 feet.

H-11 REV 35 6/02 Activation of the time history accelerographs causes visual and audible annunciation in the control room to alert the plant operator that an earthquake has occurred.

d. Fire Detection The fire-detection system at VEGP includes smoke, flame, and temperature detectors and manual fire

alarms. Fire-detection systems are provided in all

areas with safe shutdown equipment, as well as other

locations throughout the plant. In addition to

initiating fire-suppression systems, indications

from the fire-detection system are transmitted to

the control room.

2. PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM (PERMS) PERMS receives and processes radiological input readings during normal and abnormal operating and

accident conditions; measures, evaluates, and reports

radioactivity in designated areas; and monitors

releases of radioactive materials in liquid and

gaseous effluents. Data from PERMS are obtained for

the IPC. A more detailed description of PERMS is

provided in FSAR section 11.5. PERMS consists of the following components:

• PERMS Display Computer:

capable of nonvolatile mass data storage and provided with all necessary

programming, communication, display, and hardware to

serve as information manager for the PERMS.

• Communications Display Computer:

in the control room, polls individual data modules and updates

alarm status and system radiation levels throughout

the plant. Communicates with the IPC to provide

data to general plant areas.

• Safety-related display console:

in the control room, provides remote control and status display of

the safety-related channels from the safety data

modules.

H-12 REV 56 09/10

• Data processing modules (DPMs):

microprocessor-based data acquisition processors for local control and data processing; have two-way digital

communication with the communication display

computer except safety related DPMs which have one

way communications.

• Radiation monitors:

along with their associated DPMs, detect, compute, and indicate the radiation

levels at selected plant locations and actuate

alarms if these levels exceed predetermined values;

include area, airborne, and liquid monitors.

• Displays: - PERMS Display Computer. - Communications Display Computer.

• Printers: - Radiochemistry Lab.

There are four types of radiation monitors in PERMS:

area radiation monitors, airborne and air particulate radiation monitors, liquid radiation

monitors, and post-accident radiation monitors. The post-accident radiation monitors provide radiation monitoring after an accident. The

monitors are comprised of area, airborne, and air

particulate monitors. Area monitors respond to

gamma radiation photons within any energy range from

60 keV to 3 MeV. Airborne monitors are capable of

detecting and measuring radioactive gaseous effluent

concentrations with compositions ranging from fresh

equilibrium noble gas fission product mixtures to

10-day old mixtures. Power to post-accident

monitors is diesel generator backed to ensure

against interruption of monitor operation and loss

of data. 3. Integrated Plant Computer (IPC) The Integrated Plant Computer gathers, stores, and displays data used by TSC and EOF personnel to analyze plant conditions. The IPC performs this

function independently of actions in the control

room and without degrading or interfering with

control room and plant functions. The IPC consists

of workstations, printers, and associated computer hardware and software.

H-13 REV 56 09/10 The IPC serves as the primary data acquisition system for emergency response, acquiring, processing, and feeding data to the TSC and SPDS.

In addition, data links are provided to other

locations including the EOF. Overall system

unavailability is 0.01 or less. All inputs

originating from safety systems are individually

isolated at specific locations before entering the

IPC system. The IPC has the required data storage capability to meet the guidelines of NUREG-0696, which specifies 2

hours of pre-events data, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of post event

data, and 2 weeks of additional post-event data

within reduced-time resolution. The emergency response parameters provided to the IPC are listed in table H-1. All the parameters in

the IPC are available to all IPC workstations. The IPC hardware and software are protected against unauthorized manipulation of, or interference with, input signals, data processing, data storage, and

data output. This security is provided by way of key

lock devices, integral program write protection, restricted authorized personnel access, and other

administrative controls. The required IPC equipment (processor, workstations, printer, and network devices) are powered by a battery system (uninterruptible power). Meteorological information is collected by a data recorder at the meteorological tower. The

information is transmitted via microwave to the

plant. In the plant, the data is provided to the

IPC computer. The meteorological system power is

provided with a wet cell battery backup. In the event the data transmission system fails, data will be obtained by sending a person to the

meteorological shack to collect the information and

phone it back to the plant. Fire detection information is provided via visual observation of the fire alarm panel by the control

room personnel and an audible alarm. The fire alarm

panels are discussed in the Vogtle Electric

Generating Plant (VEGP) Final Safety Analysis Report (FSAR) subsection 9.5.1. Information of fire

detection will be verbally transmitted to the other

ERFs by control room personnel. Fire main header

pressure is provided by direct input to the IPC.

Seismic panels are discussed in the VEGP FSAR

subsection 3.7.4. Information on seismic motion

will be verbally transmitted to other ERFs by

control room personnel.

H-14 REV 38 05/05 The following discussion summarizes the IPC system hardware for emergency response at various locations.

a. Computer Room Hardware The computer is located next to the lower cable spreading room. This room house the main

computer for processing information. Also

located in this room is the Input/Output

equipment, printing, off-line storage, and

computer network equipment.

b. EOF Hardware The EOF is located in Birmingham, Alabama. It contains workstations and computer network equipment required for communicating with the

main IPC processors via the IPC gateway.

c. TSC Hardware The TSC is located in the power block in the vicinity of the control room. This room houses

workstations, computer network equipment and data

link equipment.

d. Control Room Hardware The control room contains SPDS workstations and other general operator use workstations.

e. System Input Room Hardware This room is used for signal tie point to the IPC equipment when required.

f. Remote Processing Units In addition to the input configuration described above, many input signals come from remote processing units (RPUs). Field data will be scanned, converted to engineering units, and transmitted to the IPC

system via seven data links from three remote

processing units. Data from train A will be

transmitted via three data links by RPUA. Data

from train B will be transmitted via three data

links by RPUB. Nonnuclear safety data will be

transmitted via a single data link by RPUN.

H-15 REV 38 05/05

g. Process Effluent Radiation Monitoring System Radiation inputs are scanned and transmitted to the IPC system via data links by PERMS.

4. Emergency Response Data System (ERDS)

In accordance with the requirements of 10 CFR 50, Appendix E, Section VI and NUREG 1394, Revision 1, means are provided to transmit critical plant

variables from the onsite computer system to the

Nuclear Regulatory Commission Operations Center via

a dedicated communications link. The installed

system consists of a computer which polls the

Integrated Plant Computer (IPC)for the required

data, assigns the appropriate data quality value, formats the data stream, and periodically transmits

the data stream over the dedicated

telecommunications circuit to the NRC. NUREG 1394, Revision 1, required submittal of a Plant Attribute Library (PAL) and Data Point Library (DPL) to the NRC. Vogtle Electric Generating Plant

submitted the original PAL and DPL information to

the NRC under a letter from C. K. McCoy to the USNRC (letter number ELV-03500). Changes to the computer

configuration or data protocols (contained in the

PAL) must be reported to the USNRC at least 30 days

prior to installing the change. Changes to the

information describing the specific computer data

points transmitted must be reported to the USNRC

within 30 days following the change. The specific

plant parameters which are designated as ERDS points

are shown in the FSAR in table 7.5.2-1. Typical

plant modifications which might affect the DPL

include: 1. Software changes which affect calculated points on the IPC.

2. Rescaling or replacement of transmitters that are scanned by the IPC, PSMS, or PERMS and are

associated with ERDS.

5. SAFETY PARAMETER DISPLAY SYSTEM The SPDS provides a display of plant parameters from which the status of operation can be assessed, in the control room and TSC. The EOF accesses SPDS via the IPC gateway. The SPDS has the following functions:

• Aids the control room operators in the rapid

detection and identification of abnormal operating

conditions.

H-16 REV 22 5/95

• Provides additional specific information to analyze and diagnose the cause of abnormal

operating conditions.

• Monitors plant response to corrective actions.

• Provides grouping of parameters to enhance the

operators' capability to assess plant status

quickly without surveying all control room

displays concurrently.

• Directs the operators' attention to other specific

confirmatory non-SPDS control room displays.

• Provides human factors engineered display formats

in simple and consistent display patterns and

codings.

• Provides display information on a real-time basis, along with validation of data.

• Provides generated selectable trend displays on a

real-time basis for monitoring reactivity control, reactor core cooling and heat removal from the

primary system, reactor coolant system integrity, radioactivity control, containment integrity, and

other selected parameters. The SPDS in the control room consists of displays of sets of concentrated parameters from which plant

safety status can be rapidly assessed. Duplicate

SPDS displays are located in the TSC and EOF to

maximize the exchange of information between these

facilities and the control room. The SPDS in each

facility is a peripheral of the IPC system. The SPDS is in operation during normal and abnormal operating conditions. The selection of parameters to be displayed on the SPDS is based on the parameters required to monitor

the critical safety functions identified by the

Westinghouse Owners Group (WOG). These parameters

will aid control room operators in determining the

safety status of the plant. The justification for

selecting these parameters is contained in the

analyses and background information generated by the

WOG to support the critical safety function

restoration guidelines. The emergency response

guidelines, which contain the critical safety

function restoration guidelines and identify the

parameters used to monitor the critical safety

functions, have been submitted to the NRC by the

WOG.

H-17 REV 34 04/02

6. POST ACCIDENT SAMPLING Liquid samples from the reactor coolant system and the containment sumps, and air samples from the containment atmosphere may be taken during accident conditions.Section I of this plan contains a more detailed description of these capabilities.

7. OTHER PROCESS PARAMETERS Several other process parameters, including reactor coolant system pressure and temperature, containment pressure and temperature, liquid levels and other

system indications, are useful both for the

initiation phase and continued assessment. The

specific parameters monitored by the IPC are listed

in table H-1. Several of these are used in the

classification process as discussed in section D, Emergency Classification System. H.5 OUT OF PLANT MONITORING

1. GEOPHYSICAL MONITORING A source of meteorological data is Bush Field in Augusta. The National Weather Service (NWS) maintains an automated observation station at the

airport; and windspeed, wind direction, cloud cover, and ceiling height can be obtained. Information

from this automated observation station as well as

forecast information can be obtained from the NWS in

Columbia, S. C.

2. RADIOLOGICAL MONITORING VEGP will have sufficient portable equipment and trained personnel to field three field monitoring teams. Each team will include two people who will

obtain an emergency monitoring kit. The kits will

include dosimeters, a two-way radio, meters for

measuring gamma and beta/gamma dose rates, and air

samplers for collecting particulates and iodines.

The particulate filter is used in the field

primarily to clean the sample so that any activity

on the cartridge (silver zeolite or equivalent) will

be iodine. The cartridge is then counted in the

field to provide an estimate of airborne iodine

concentration. VEGP monitoring teams will remain on

the Georgia side of the Savannah River.

H-18 REV 57 06/12 Radiological monitoring on the South Carolina side of the Savannah River will be conducted by personnel from the SRS, or the State of South Carolina. These

field monitoring teams will be equipped with

equipment similar to that used by the VEGP teams.

Results of the offsite monitoring activities will be provided to the TSC until the dose assessment activities are transferred from the TSC to the EOF.

3. LABORATORY FACILITIES VEGP has laboratory facilities for analysis of radioactive samples. The major pieces of equipment include a solid-state gamma spectrometer and a

beta/gamma gas proportional counter. The GPC environmental laboratory located in Smyrna, GA has the capability to perform isotopic analyses

of drinking water, river water, milk, vegetation, sediment, and biological samples, as well as tritium

and gross-beta analysis. In addition, processing of

environmental optically stimulated luminescent dosimeters (OSLDs) will be handled by this laboratory. Backup laboratory facilities are available at Plant Hatch. This backup capability could be used if

facilities in VEGP were not available. H.6 EMERGENCY KITS Emergency kits are located in the TSC (for control room also), the OSC, the health physics control point, the

EOF and other plant locations. An ambulance kit will be

carried by the VEGP health physics technician who

accompanies the ambulance. Procedures require an

inspection and operational check of equipment in these

kits on a quarterly basis and after each use. Equipment

in these kits is calibrated in accordance with the

suppliers' recommendations. A set of spares of certain

equipment is also maintained to replace inoperative or

out-of-calibration equipment.

A listing of the typical contents of each kit and the spares is included in appendix 4.

REV 34 4/02 TABLE H-1 INTEGRATED PLANT COMPUTER INPUTS

• Neutron flux

• Pressurizer (pressure, level, temperature, valve position).

• Reactor coolant system (pressure, temperature, flow)

• Reactor coolant pump status

• Containment (pressure, temperature, H2 concentration, water level)

• Containment spray system status

• Safety injection system status

• Residual heat removal (RHR) system status

• Chemical and volume control system (CVCS) status

• Nuclear service cooling water (NSCW) system status

• Component cooling water (CCW) system status

• Auxiliary component cooling water (ACCW) system status

• Steam generator (pressure, level, flow)

• Auxiliary feedwater system status

• Turbine status

• Generator status

• Condensate system status

• Feedwater system status

• Circulating water system status

• HVAC systems status

• Radiation monitoring systems (effluent, area, and process

monitors)

• Meteorological system (primary and backup tower, wind speed, wind direction, indication of stability class, temperature) {Unit One IPC only}

VESTIBULEAIRLOCKCOMPUTERROOMAIRLOCKMECHANICALEQUIPMENT ROOM(HVAC)COMMUNICATIONSCONFERENCE ROOMELECTRICEQUIPMENTROOMBATTERYROOMRESTROOMS& SHOWERSWORK AREA LEVEL 1CONTROLBUILDINGAIR LOCKAIR LOCK TOCONTROL ROOM(ERDS COMPUTER)(IPC TERMINAL)

• • • • • • • • OPERATION ROOMEMERGENCY DIRECTOR TSC MANAGER ENGINEERING SUPERVISOROPERATIONS SUPERVISORMAINTENANCE SUPERVISORHEALTH PHYSICS SUPERVISORCHEMISTRY SUPERVISORNRC DIRECTOR ( IPC TERMINALS)

• • NRC STAFF VEGP STAFF*HP/CHEM STATT (IPC TERMINAL)65'65'84'TECHNICAL SUPPORT CENTERVogtle Electric Generating PlantFIGURE H-1TSC LAYOUT PLANREV 34 4/02

OSC STAFFCONFERENCEAREAOPERATIONS SUPPORT CENTEROSC STAFFWAITING AREAOSCMANAGER'SOFFICE S T O R A G ESTAIRWELLMAINTENANCE BUILDINGSECOND FLOORMAINTENANCEENGINEERINGOFFICESKITCHENWOMEN'SRESTROOMSTAIRWELL NVogtle Electric Generating PlantFIGURE H-2REV 57 06/12OSC LAYOUT PLANMAINTENANCEOFFICES I-1 REV 56 09/10 I. ACCIDENT ASSESSMENT This section describes the methods, systems, and equipment available for assessing and monitoring actual or potential offsite

consequences of a radiological emergency. Initial assessment

actions are the responsibility of the shift superintendent and/or

the shift supervisor, using available shift personnel. Subsequent

assessment actions are directed by the emergency director with

assistance from the control room, technical support center (TSC),

emergency operations facility (EOF), and emergency teams, as

necessary.

I.1 PLANT PARAMETERS Plant system and effluent parameter values characteristic of the spectrum of off-normal conditions and accidents and the manner in which these values are used to classify an

emergency are provided in section D. Some of the parameters

monitored include reactor coolant system pressure, reactor/pressurizer water level, containment pressure, containment radiation level, effluent monitor readings, and

area radiation monitor readings. NMP-EP-110, "Emergency Classification Determination and Initial Actions, includes methods for quickly assessing plant system and effluent

parameter values and classifying the emergency condition.

Additional information on plant instrumentation is provided

in section H.

I.2 RADIOLOGICAL MONITORS In-plant radiological measurements provide information to help assess emergency conditions. The containment high-range radiation monitor and containment hydrogen monitor are used

to provide an early indication of the quantity of

radioactivity available for release from the containment.

Emergency procedures (Procedure 91502-C) include a

correlation between the monitor reading and the extent of

core damage. These correlations are based on the

Westinghouse Owners Group (WOG) Core Damage Assessment

Guidance, WCAP-14696-A, Revision 1, November 1999. A more

detailed assessment of core damage is then performed using

Procedures 91502-C and 91503-C, which are also based on the

WOG methodology.

I-2 REV 57 06/12 Data required to evaluate core conditions and coolant chemistry conditions are obtained through established chemistry procedures. Samples can be obtained for reactor

coolant, containment sump, and containment atmosphere and are

used for all radiochemical analyses.

Reactor coolant grab samples can be either diluted or undiluted. Grab samples can be transported in a shielded

cask. Analysis capability meets all Final Safety Analysis

Report (FSAR) commitments to Regulatory Guide 0737, II.B.3.

In addition to the onsite capabilities for radiological

assessment, AREVA NP has agreed to provide backup analysis of high radioactivity level samples.

The TSC manager will approve and direct the transport of the grab sample to AREVA NP. A transport cask will be obtained from the Pooled Inventory Management Systems (PIMS).

Chemistry personnel will collect the sample in the sampling

cask and transport it to the loading area. Documentation

will be completed and the transport cask shipped to AREVA NP.

I-3 REV 64 3/15 I.3 DETERMINATION OF RELEASE RATE The source term or release rate is determined using the process and effluent radiation monitoring system (PERMS) and measured or estimated flowrates for releases via monitored effluent

release paths (plant vent, turbine building vent, or steam

generator relief valves) per procedure NMP-EP-104, Dose Assessment.

In the event that instrumentation is off-scale or inoperative, direct measurements with portable survey instruments will be

used for determination and verified by field monitoring team

samples.

I.4 DOSE ASSESSMENT SYSTEM Computer dose calculation systems will be located in both the TSC and EOF for offsite dose assessment purposes. These systems will support the MIDAS code, a Vogtle-specific version

of a dose assessment computer code developed by Pickard, Lowe, and Garrick, Inc., (PLG). The code is based on a program

written by PLG called CRACIT (Calculation of Reactor Accident

Consequences Including Trajectory), which is similar in concept

to that in the CRAC program written for the Reactor Safety

Study. The MIDAS version is referred to as MIDRAC. The basic

functions of MIDRAC are the calculation of dispersion of the

released material as it travels downwind and the estimation of

the resulting concentrations of this material. Dispersion is

modeled using the straight-line Gaussian dispersion model and

variable trajectory transport. Initial dose projections can be

made within 15 min of a radiological release utilizing the

computer system. Subsequent dose projections will be made

approximately every 15-30 min depending on the variability of

meteorological conditions and/or radioactive releases.

I-4 REV 25 04/97 MIDAS is a personal computer based program for rapidly assessing the radiological impact of accidents at nuclear

power plants. It calculates total effective dose equivalent (TEDE), thyroid doses, and skin doses at various fixed

downwind distances. Source term information is derived from

plant effluent monitors, RCS or containment samples, field

monitoring teams, or default accident scenario.

Actual meteorological data and release rate data are obtained from the plant computer and entered into the dose projection

computer. Minimum meteorological data to be obtained include

windspeed, wind direction, and a stability indicator (either

vertical temperature difference or standard deviation of the

horizontal wind direction). Plant-specific default values

are part of the program for use when meteorological or

release rate data are not available. The computer will

calculate dispersion, dose, and plume arrival times. Dose

calculations are based on dose conversion factors (DCF) from

EPA 400.

Default release rates are available for possible accidents in the event that measured source term data are not available or

the case where bounding calculations are desired. Default

values for various accident source terms are presented in

chapter 15 of the FSAR.

Meteorological data are obtained and utilized as input to the dose model, as described in section H.4.1.a. Data from the

primary meteorological monitoring system can be accessed

directly from the control room, TSC, and EOF, and are also

available to NRC personnel and State representatives at the

site. In the event the primary instruments are unavailable, the backup meteorological tower is equipped with instruments

at the 10-m level to provide parameters relevant to

atmospheric dispersion calculations (i.e., windspeed, wind

direction, and sigma theta). In the event both the primary

and backup meteorological systems are unavailable, meteorological data will be obtained by commercial telephone

directly from the National Weather Service located in Columbia, South Carolina. NWS Columbia has access to information from the automated weather station at Bush Field, Augusta, Georgia. Forecast changes in wind direction will be utilized in determination of expected changes in plume

trajectory. These forecast changes in plume trajectory may

be utilized to expand the areas for which protective actions

are recommended.

In the event that significant wind speed or stability class changes are expected, the effect of the expected changes on

dose projections will be analyzed utilizing the dose

assessment model.

I-5 REV 38 06/04 In cases where weather forecasts predict precipitation, this information will be utilized in reference to adverse weather evacuation time estimates as appropriate. When precipitation

is predicted or occurring in the area of the plume, the

potential for significantly increased rates of radioactivity

deposition will be considered by increasing the scope of

environmental sampling as required in order to quantify the

effects of this potentially increased deposition.

The VEGP staff will calculate the 50-mile ingestion pathway doses from the deposition of specific radionuclides.

The VEGP field monitoring team will collect sufficient environmental data to characterize the initial deposition of

activity, the peak activity in pasture grass and milk, and

total intake of I-131, Cs-137, Sr-90, and Sr-89. The samples

will be analyzed at the VEGP site, the Environmental

Laboratory in Smyrna, Georgia, or at the Plant Hatch

Laboratory. The analysis results will be compared with the

Preventive and Emergency PAGs and the associated doses will

be determined.

The dose assessment computer program will be used to calculate the projected deposition of radionuclides and

associated doses in the ingestion pathway based on release

data and meteorological conditions. These estimates will be

compared to the Preventive and Emergency PAGs. The results

of all analyses will be provided to the States of Georgia and

South Carolina by the dose assessment supervisor.

Each state is responsible for implementing protective measures based on protective action guides and other

criteria. The measures will be consistent with the

recommendations of HEW/FDA regarding contamination of human

food and animal feed.

I.5 FIELD MONITORING The emergency director or his designee can deploy up to three teams for field monitoring. These teams are available for field monitoring within the plume EPZ as described in section

H. Initially, the emergency director can activate at least

one team from on-shift personnel. Once the emergency

facilities are activated, the emergency director can request additional monitoring teams from support personnel located

at the operations support center (OSC).

Monitoring teams of at least two people are formed and dispatched at the OSC or TSC as appropriate. Field monitoring kits are available to the teams in predetermined locations as described in 91702-C. The kits will include anti-contamination suits, respirators, a two-way radio, meters for measuring gamma and beta/gamma dose rates, and air

samplers.

I-6 REV 61 06/14 Designated vehicles may be used which are equipped with two-way radios on plant-dedicated frequencies. Handheld radios

will also be available as a backup. Vehicles will be

available 24 h per day.

Prior to leaving for the field, the dose assessment supervisor, or designee, will normally direct and brief the

teams on the initial survey and sample locations, suggested

travel routes, meteorological conditions, and team

identification name or number for communication purposes.

The teams will inspect their field monitoring kits, perform

survey equipment operation checks, obtain dosimeters, and

establish radio communications with the monitoring team

communicator. Monitoring teams are instructed to contact the

monitoring team communicator approximately every half hour.

The teams will obtain their briefing from the health physics supervisor in the TSC or in the OSC by the OSC manager. They

will be controlled by the TSC until the responsibility is

transferred to the EOF.

It is estimated that teams will be in the field and performing monitoring tasks within about 1 h of the

determination of the need for field monitoring.

Additional field monitoring team formation and dispatch details are contained in Procedures 91201-C, 91202-C, NMP-EP-102, "EPF Manager," NMP-EP-104, "Dose Assessment Supervisor," NMP-EP-101, "Activation and Operation of the EOF," and NMP-EP-106, "EOF Support Coordinator."

Preselected radiological sampling and monitoring locations are designated in implementing procedures and are shown on

figure iii. Field monitoring teams may be directed to

perform sampling at these locations and others by the dose

assessment staff at the TSC or EOF. In-transit dose rate

measurements will be made. The teams may conduct airborne

and dose rate measurements near the expected plume

centerline.

If the dose rate exceeds 100 mrem/h, off-centerline measurements will be made. Based on dose rates, the teams

will be directed to sweep the plume to identify the

centerline or maximum dose rate. Dose rate readings will be

taken with open and closed window on the radiation meter.

When the ratio between open and closed window readings is a

factor of 2 or higher, it will be assumed that the

measurement was taken in the plume. If the open and closed

window readings are the same, it indicates that the plume is

probably overhead.

I-7 REV 38 06/04 The emergency monitoring kits contain a portable air sampler, silver zeolite cartridges, and counters to provide the capability to detect and measure radioiodine concentrations

in the air as low as 10

-7 µCi/cc. The list of equipment carried by the field teams is described in appendix 4, Table

4-3. Procedure 91303-C, describes the sampling and measuring

techniques for air samples.

The total sample volume and the limiting background count rate allow for a LLD of at least 10

-7 µCi/cc. The cartridges can be counted in the field without interference from noble

gas (background count rate below 300 cpm on an HP-210 probe

or equivalent). The cartridge and air particulate filter

will be returned to the laboratory at the plant for isotopic

analysis if the field analysis reading is 100 cpm above

background on an HP-210 probe or equivalent.

Depending on wind direction and/or the severity of the incident, additional field monitoring teams may be provided

by DNR, SC-DHEC, DOE-SR or other divisions of the DOE.

Coordination of these teams and data transfer will be

accomplished using existing communication links (see Section

F for details). The state and VEGP field monitoring teams

will be coordinated by the dose assessment supervisor to assure a fully coordinated effort. DOE-SR will direct the

field monitoring teams of the Savannah River Site depending

upon the wind direction. DOE-SR will make their monitoring

data available to VEGP and State and local representatives at

the EOF. The Dose Assessment Team at the EOF will collate

field monitoring data for VEGP dose projection purposes.

This information will be available to the State and local

representatives at the EOF and to DOE-SR.

I-8 REV 57 06/12 I.6 ENVIRONMENTAL SAMPLES In addition to direct monitoring and air sampling, the assessment program includes an emergency environmental sampling program in which routine types of environmental

samples (water, air, soil, and vegetation) are collected and

analyzed in the laboratory for detailed radionuclide data. The GPC environmental laboratory, located in Smyrna, Georgia, has the capability to perform isotopic analyses of drinking

water, river water, milk, vegetation, sediment, and

biological samples as well as tritium and gross beta

analysis. A detailed description of the current fixed environmental monitoring program, which is summarized below, is presented

in the Plant Vogtle Offsite Dose Calculation Manual. Fixed

environmental sampling and monitoring locations are described

in Procedure 91303-C and are shown on figure iii. Copies of

the map showing the locations will be in the field monitoring

kits, the TSC and EOF. This program may be modified as per

VEGP Technical Specifications. Direct radiation is measured by optically stimulated luminescent dosimeters (OSLDs) which are exchanged and analyzed (read) for gamma dose quarterly. Airborne radioiodine and particulates are collected at control and indicator stations by continuously drawing a

sample stream of air through a charcoal cartridge and a

particulate filter using an air displacement pump. The

cartridge and filter are changed weekly. The cartridge is

analyzed weekly for I-131; the filter is analyzed for gross

beta following filter change. Filters are composited

quarterly by location for gamma isotopic analysis. Surface water samples are taken from the Savannah River by continuous automatic sampling equipment and collected in

containers resulting in composite samples for each location.

Composite samples are collected monthly and analyzed for

gamma isotopics. Quarterly composites, which are obtained

from the monthly composites, are analyzed for tritium. Samples from drinking water supplies are taken by continuous automatic sampling equipment near the intakes of water

treatment plants. Composite samples from each location are

collected monthly. Grab samples of finished drinking water

are taken monthly. Monthly samples are analyzed for gross

beta and gamma isotopics. Quarterly composites, which are

prepared from the monthly samples, are analyzed for tritium. Sediment from the Savannah River is collected semiannually from control and indicator locations using manual grab

sampling techniques. Sediment samples are analyzed for gamma

isotopics.

I-9 REV 57 06/12 Milk samples are collected semimonthly by taking grab samples from milk supplies at control and indicator stations. These samples are analyzed for gamma isotopics. Vegetation is sampled monthly during the growing season by collecting grab samples of grass or leafy vegetation at

control and indicator stations. Vegetation samples are

analyzed for gamma isotopics. The normal environmental sample analysis is performed at the Georgia Power Company's Environmental Laboratory located in

Smyrna, Georgia. During and/or subsequent to emergency

conditions, the routine environmental monitoring program will

be modified to collect and analyze additional samples from

existing stations. The dose assessment supervisor will

coordinate sampling and analysis activities for those areas

that may have been impacted by a release from the plant.

Sample results will be transmitted back to the dose

assessment supervisor by the analyzing organization. Data from fixed monitoring stations (OSLDs and air samplers) will be utilized to estimate population dose. The samples

from fixed monitoring stations would be collected after

termination of a radioactive release and analyzed. The

results would then be reduced in a manner that would assist

in defining the trajectory, radioactivity, and impact of the

released plume.

I.7 USE OF FIELD MONITORING DATA Field monitoring measurements are important in determining the actual radiation levels in the environment. The dose assessment computer code provides only a rough approximation

of radiation levels and location of the plume. The

uncertainties in the source term and meteorological

conditions in the affected areas are the chief contributors

to the inaccuracies of projected dose and dose rate. Once

field monitoring information becomes available, the dose

assessment computer code becomes less important in providing

data on which to base protective action recommendations.

There is no widely accepted formula to use field monitoring

data to reduce the uncertainties and inaccuracies in the dose

assessment computer code. The dose assessment supervisor or

TSC HP supervisor must exercise professional judgement in

determining the proper correction factors. Field monitoring teams should not unnecessarily be exposed to high levels of radiation from a radioactive plume. ALARA

principles should be applied to all field monitoring

activities. Once the decision to evacuate the general public

has been made, field monitoring activities should be

restricted to such activities as determining the source term

for an unmonitored release or confirming the presence of a radioactive plume. Plume centerline tracking is of

I-10 REV 21 10/94 relatively minor importance in protecting the health and safety of the public and should not be attempted for highly radioactive plumes unless the information derived has some value in assisting in the evacuation of the public. Field monitoring team's activities that contribute to the expeditious evacuation and consequent reduction in radiation dose of the public should continue within the EPA emergency exposure guidelines if necessary.

J-1 REV 35 6/02 J. PROTECTIVE RESPONSE This section describes the protective actions that have been developed to limit radiation exposure of plant personnel and

the public following an accident at the plant. This section

addresses conditions at the Alert, Site Area, or General

Emergency classification. Protective response at the

Notification of Unusual Event (NUE) level would be taken at

the discretion of the emergency director.

J.1 PROTECTIVE RESPONSE FOR ONSITE PERSONNEL Protective response for onsite personnel (including visitors and contractor personnel) depends upon alerting, assembly and accountability, site dismissal, monitoring, and decontamination.

1. ALERTING Section E of this Plan, Notification Methods and Procedures, describes the methods to be used to alert onsite personnel of emergency conditions.

2. ASSEMBLY AND ACCOUNTABILITY A security related emergency may preclude the ordering of assembly and accountability in order to protect plant personnel from the security threat.

The decision not to order assembly and accountability

will be made by the Emergency Director. Upon activation of the plant emergency alarm, plant personnel assigned specific emergency

responsibilities proceed to their designated

emergency response location. Emergency response

personnel in the protected area badge into their

emergency response facility (TSC, OSC, or control

room) using their ACAD badge and also sign in on a

personnel roster (TSC and OSC only). The security

computer system performs an initial accountability of

all persons in the protected area. Thereafter, the

emergency response facility managers of the control

room, TSC, and operations support center shall be

responsible for periodically assuring that

accountabilities in their facilities are being

maintained. Assignment logs and required periodic

communications between emergency response teams

maintain accountability. Non-involved plant personnel, visitors, and contractors located within the protected area leave

the protected area upon hearing the emergency alarm

and report to their designated assembly areas. As

these individuals exit the protected area, they badge

out. The Security Department accounts for each

person inside the protected area at the start of an

emergency by using the security computer system.

J-2 REV 39 11/04 This method provides for accountability of all individuals inside the protected area within about 30 min. of the emergency declaration page

announcement. Accountability reports are made

periodically to the emergency director by the

Security Department.

3. SEARCH AND RESCUE If protected area accountability reveals a missing person, the emergency director assembles a search and rescue team per emergency response procedures (Procedure 91401-C, Assembly and Accountability).

The search and rescue team obtains information on

last known location from the computer system or

reports from other personnel. A search of likely

areas is conducted until the missing individual is

located.

4. SITE DISMISSAL Site dismissal, with or without monitoring, of non-involved personnel on-site (if feasible) is ordered by the emergency director whenever a Site Area or

General Emergency is declared. If there has been no radioactive release and a release is not projected, the emergency director may

elect to order a "site dismissal with no monitoring" rather than with monitoring. For a site dismissal

with no monitoring, non-involved personnel are sent

home without going to reception centers. If site dismissal with monitoring is necessary, the emergency director will notify Burke County EMA and request setup of a reception center to receive VEGP non-involved personnel. The route selected to the reception center is based on meteorological and/or radiological conditions. The location of the reception center is shown in figure iv in the Preface Personnel on site will be notified by public address, site siren, or other communication

that dismissal of non-involved personnel to the

applicable reception center will take place and specify the route. The site dismissal will be

conducted and controlled per Procedure 91403-C, Site Dismissal. Security will dispatch security officers

to work areas outside the protected area to ensure

all non-involved personnel have left the owner-

controlled area.

J-3 REV 61 06/14 Upon site dismissal to a reception center, non-involved personnel will be monitored for contamination to determine gross contamination as

per the Burke County Emergency Operations Plan.

Those personnel who are contaminated will undergo a

decontamination process in accordance with standard

health physics procedures. Those personnel who are

not contaminated will be released upon clearance of

the vehicle. Vehicles will be monitored by gamma instruments for contamination in the designated parking areas.

Those vehicles which indicate contamination will be

marked or identified for decontamination.

Uncontaminated vehicles will be allowed to exit the

area upon authorization of the reception center

emergency workers. Contaminated vehicles will be

decontaminated as per the Burke County Emergency

Operations Plan. Contaminated articles and clothing and waste material will be collected and placed in containers

or bags for disposal and/or processing at the site.

5. SECURITY EVENTS Onsite protection of employees during hostile action involves a combination of restricted movement, movement to safe locations, and site evacuation depending on the nature of the hostile event and advance warning. Site procedures provide specific actions to take during hostile action based events.

These actions will be communicated to onsite personnel via the plant PA system and other communications means as applicable.

6. MONITORING AND DECONTAMINATION When an Alert is declared and site dismissal with no monitoring is anticipated, personnel who have left the protected area are monitored by portal monitors.

If necessary, decontamination is completed using the

plant decontamination facilities located in the

control building or other appropriate location. When site dismissal with monitoring is expected and release of radioactivity has occurred, monitoring is

performed by Burke County emergency workers at an

established reception center.

J-4 REV 38 06/04 Should decontamination be necessary, the reception center establishes a field decontamination area, using materials from emergency kits located in the

vicinity of the reception center. Decontamination

and waste disposal are completed in accordance with

the Burke County Emergency Operations Plan.

6. USE OF ONSITE PROTECTIVE EQUIPMENT AND SUPPLIES A supply of potassium iodide is stored in the TSC for TSC and control room use, OSC, main control point, or health physics room. The health physics supervisor will direct the issuance of potassium

iodide when the projected thyroid exposure is

greater than 25 rem. The health physics supervisor

will direct radiological survey personnel to

distribute potassium iodide and record the name and

social security number of those individuals who are

issued potassium iodide. Potassium iodide will be

issued in 130-mg doses daily for at least 3 days, but not more than 10 days. Issuance will be

performed immediately prior to exposure or not

longer than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after exposure. At the time potassium iodide is distributed, an iodine sensitivity check will be made by querying each

individual concerning known reactions to iodine.

Individuals who have experienced reactions to iodine

will be excused from duties requiring issuance of KI.

J.2 PROTECTIVE RESPONSE FOR THE PUBLIC VEGP is responsible for ensuring that timely recommendations for protective actions reach appropriate State and local officials. These officials (as described

in section A) are then responsible for alerting the

public and ordering shelter and/or evacuation, if

necessary.

1. ALERTING The means used by VEGP to alert local and State agencies and the means used by local and State agencies to alert the public are described in

section E and appendix 3 of this Plan.

J-5 REV 63 10/14

2. PROTECTIVE ACTION RECOMMENDATIONS The emergency director is responsible for providing protective action recommendations to State and local officials as part of initial notifications and

follow-up communications of a General Emergency.

These recommendations are based upon assessment

actions as described in section I of this Plan.

Using available information on plant conditions, projected dose estimates, and any available

monitoring data, the emergency director recommends

whether the public should be advised to seek shelter

or evacuate based on the Environmental Protection Agency (EPA) Protective Action Guidelines.

Procedure NMP-EP-112, Protective Action Recommendations, provides detailed guidance on protective action recommendations (PARs) which were developed in accordance with NUREG-0654, Supplement 3, Rev. 1. Offsite Response Organizations (OROs) were actively involved in the development of the PARs. The emergency director will approve the PAR decision developed using the methodology from the Flowcharts shown as Figures J-1, Initial and J-2, Follow-up. The mechanism for communicating these recommendations is described in section E of this

Plan. 3. EVACUATION AND SHELTERING The Georgia Emergency Management Agency (in coordination with Georgia Department of Natural Resources) and South Carolina Emergency Management

Division (in coordination with South Carolina

Department of Health and Environmental Control) are

responsible for deciding protective measures for

affected offsite areas within their jurisdictions.

State officials will consider the potential risks of

implementing protective actions against the

reduction of radiological risk achieved by the

protective action. Determination of the benefit of evacuation must take into account the time needed to complete the

evacuation. Appendix 6 includes more detail on how

these estimates were developed and presents

information on evacuation routes, evacuation areas, relocation centers, shelter areas, and the

population distribution by evacuation areas and

sectors.

J-6 REV 11 1/89 If a decision is made to evacuate any part or all of the plume exposure pathway EPZ, the evacuation will be carried out in accordance with the emergency

response plan of each affected county. In the event of an evacuation, the populace will be instructed to proceed by the appropriate evacuation

route to predesignated reception centers/shelters. Reception centers/shelters for Georgia and South Carolina counties within the plume exposure pathway

EPZ are listed in Table J-5. The services to be

provided in the reception centers include:

• Registration

• Screening for contamination

• Decontamination as needed

• Information and assistance for family unification

• Food and lodging

• First aid Privately owned vehicles will be the primary mode of transportation if evacuation is directed.

Individuals who do not have their own means of

transportation have been advised in the public

information calendar, to arrange their own transportation, if possible. If this is not

possible, individuals are instructed to stay tuned

to the radio or television and listen for the phone

number to call to be picked up. Specially equipped

vehicles will be dispatched directly to the homes of

handicapped and/or nonambulatory individuals

requiring special transportation means. Under certain conditions, sheltering inside the home may be the preferred recommended action. Area radio

and television stations or tone alert radios will

advise the public on taking this action, will

provide instructions to the public, and will give

the "all clear signal" when appropriate.

REV 22 5/95 TABLE J-1 USE OF PROTECTIVE EQUIPMENT AND SUPPLIES Equipment Criteria for Issuance Location Means of Distribution Full face canister respirator As needed by onsite

emergency teams in

areas of high airborne

radioactivity a. Emergency kits

b. Health physics (HP) supply room a. Issued at OSC or control point b. Issued as needed by HP personnel Self contained breathing apparatus a. Firefighting

b. Toxic Gas

c. Highly radioactive airborne activity d. Lack of oxygen a. Control room

b. Emergency kits a. Used as needed by operators Protective clothing (coveralls, hoods, boots, gloves)

As needed in areas of

known contamination a. Various areas of the station b. Emergency kits

c. HP supply room a. Issued as needed by HP personnel b. Issued at OSC or control point for

emergency teams

REV 63 10/14 TABLE J-2 Table J-2 has been deleted.

REV 63 10/14 TABLE J-3 Table J-3 has been deleted.

REV 61 06/14 TABLE J-4 EVACUATION TIME ESTIMATE

SUMMARY

Table J-4 has been Deleted.

REV 39 11/04 TABLE J-5 (SHEET 1 OF 2)

RECEPTION CENTERS/SHELTERS FOR COUNTIES WITHIN PLUME EXPOSURE PATHWAY EPZ OF VEGP COUNTY RECEPTION CENTER/SHELTER CAPACITY 1. Georgia Burke County Burke County Comprehensive 4,675, school High School in session 1057 Perimeter Road Waynesboro 5,980, school not in session

2. South Carolina Aiken County PRIMARY South Aiken High School 1200 701 Pine Road Aiken, SC 29801 (803) 641-2600 ADDITIONAL CENTER Kennedy Middle School 1200 659 Pine Log Road Aiken, SC 29801 (803) 641-2470 Middlebrook Elementary School 255 East Pine Log Road Aiken, SC 29801 (803) 641-2580

REV 41 12/05 TABLE J-5 (SHEET 2 OF 2)

COUNTY RECEPTION CENTER/SHELTER CAPACITY Allendale PRIMARY County(a) Allendale-Fairfax High School 3581 Allendale-Fairfax Highway (278 E.)

Allendale, South Carolina 29810 (803) 584-2311

675 ADDITIONAL CENTER Allendale Elementary School 4561 Allendale-Fairfax Highway (278 E.)

Allendale, South Carolina 29810 (803) 584-3476 400 Barnwell See Allendale County County Shelters listed above

a. To accommodate residents of Barnwell County within the Vogtle EPZ.

K-1 REV 57 06/12 K. RADIOLOGICAL EXPOSURE CONTROL K.1 EMERGENCY EXPOSURE GUIDELINES During an emergency, it may be necessary to authorize radiation exposures above 10 CFR 20 limits. These higher exposures may be necessary to complete protective, corrective, or lifesaving actions. Table K-1 presents

the emergency exposure limits for emergency workers

involved in protecting valuable property, protection of

large populations, or lifesaving actions. Under all such

situations, every reasonable effort will be made to

minimize exposures. Decisions as to appropriate

exposures, considering the action required and relative

risks, will be made by the emergency director in

consultation with health physics personnel. Equipment and facilities have been designed in accordance with Title 10, Code of Federal Regulations, Part 50 (10

CFR 50), Appendix A, General Design Criteria for Nuclear

Power Plants. Criterion 61, "Fuel Storage and Handling

and Radioactivity Control," requires systems which may

contain radioactivity to be designed to assure adequate

safety under normal and postulated accident conditions.

Plant design has undergone an extensive As Low As

Reasonably Achievable (ALARA) review. The ALARA reviews

ensured that the design philosophies established in

Regulatory Guide 8.8 were considered at the design stage.

Design features are considered for potential exposure and

changes are recommended to reduce potentially high doses. The post-accident sampling procedures have been designed to provide adequate protection to personnel during the

collection of grab samples. Designated sample points are

specified in plant procedures. A plant shielding design review was conducted in accordance with the criteria for infrequently occupied

areas in NUREG 0737, Item II.B.2. The projected dose

rates in the facility are presented in table 12.3.1-5 of

the VEGP FSAR. The effluent sampling procedures have

been written to assure that no individual receives a

dose in excess of regulatory criteria as designated in Table K-1. A time and dose rate study has been conducted to assure that the exposure criteria can be

achieved under accident conditions.

K-2 REV 21 10/94 K.2 ONSITE RADIATION PROTECTION PROGRAM When necessary, the emergency director can authorize emergency exposures in excess of 10 CFR 20 limits but within the limits in table K-1. Where possible, the

normal radiation work permit (RWP) procedure will be

used to control exposures. This procedure requires

signature approval, prior knowledge of worker past

exposures, and guidance on protective actions to be used

in the course of the emergency work. If time and

urgency do not allow this procedure to be followed, the

health physics supervisor may approve emergency RWP

controls. In all cases, a briefing is given to the

emergency team by health physics staff, and each team is

accompanied by a qualified health physics technician who

meets the qualifications of ANSI 18.1 criteria set out

in the VEGP Technical Specifications. This briefing

includes a discussion of the hazards involved in the

planned action, as well as protective actions to be

taken.

A record of individual and collective exposure received during the emergency will be maintained by the dosimetry

team. Exposure records at the control point or the OSC

will be updated after each entry into a radiologically

controlled area. This may be accomplished through the

dosimetry records computer system or manually. An

individual's dose margin will be assessed by determining

the difference between the updated exposure and current

administrative limit; these margins are used to

determine emergency assignments. Operation of the

manual system and activation of the dosimetry team are

described in Procedure 91110, Duties of the Health

Physics Supervisor (TSC).

In situations where exposures in excess of 10 CFR 20 limits are authorized, the following considerations will

be made prior to emergency team selection:

1. Declared pregnant female employees shall not be allowed to participate.

2. For doses greater than 25 rem, personnel shall be volunteers and be fully aware of the risks involved. All emergency exposures will be included in personnel radiation exposure records.

Emergency dosimetry is provided to each member of the emergency response organization for both onsite and

offsite organizations as required by the radiological

conditions existing at the time. Appendix 4 presents

information on the types of dosimetry available in each

emergency response facility and other locations.

K-3 REV 57 06/12 Emergency response personnel will be made aware that self reading dosimeters should be checked every 15 to 30 min. during the emergency. There is the capability to

read optically stimulated luminescent dosimeters (OSLDs) within 24 h. They will also be read if the individual

has received greater than a previously established value

as determined by health physics procedures on their

direct reading dosimeter.

Radiation dose will normally be controlled by the health physics supervisor within the limits authorized by

routine station health physics procedures. The 10 CFR

20 limits will not be exceeded without the prior

approval of the emergency director.

OSLDs are processed on a routine basis as delineated in chapter 12 of the FSAR.

K.3 DECONTAMINATION The action levels for determining the need for decontamination of personnel, equipment, and areas are delineated in plant admin. and health physics

procedures. Personnel decontamination facilities are

located on Level 1 of the control building in room

R-110. The decontamination facility is located adjacent

to the health physics station. Instrumentation to

survey personnel during and after decontamination is

located at the health physics station. The facility has

vertical showering and normal wash sinks.

Decontamination equipment for personnel is similar to

that available in the Decontamination Emergency

Equipment Kit (Appendix 4), except that the available

supply is greater and stronger cleaning solutions are

available. Waste generated through the use of the

decontamination facilities is collected and processed by

the plant liquid radwaste system.

Decontamination of personnel will be conducted in accordance with standard health physics practices.

If decontamination activities are required, a controlled access area will be established by roping off the area.

Procedure 91306, Contamination Monitoring and

Decontamination, addresses field decontamination and

waste control. Supplies of clean clothing will be made

available. Personnel decontamination will be

accomplished using water washes or other methods for

extreme cases as described in plant health physics

procedures. These procedures will be applicable to

removal of radioiodine from the skin. Decontamination

of serious wounds will be accomplished at Doctors

Hospital or the Burke Medical Center as described in

section L of this Plan.

K-4 REV 23 4/96 Equipment and area decontamination will be conducted as determined by the TSC manager, maintenance supervisor, operations supervisor, or health physics supervisor. It

is accomplished as described in plant health physics

procedures and ranges from vacuum cleaning to wash downs

with water and acid or caustic solutions.

Personnel exiting the radiation-controlled area will be monitored for contamination by stand-up monitoring booths

or by a whole-body scan with a hand-held probe. The standard health physics contamination limits will be used for release of personnel. The decontamination facilities described above can accommodate both men and women who

indicate low and high levels of contamination. Plant

areas that require access to facilitate recovery

operations will be surveyed with portable instruments

equipped with Beta/Gamma detectors. Appropriate

protective clothing will be worn, as determined by this

survey, to perform activities in these areas. Recovery

operations will necessitate more detailed surveys on an

as-needed basis.

K.4 ONSITE RADIOLOGICAL CONTAMINATION CONTROL Access control is provided by the Security Department during emergency conditions. Only authorized emergency response personnel are allowed to enter the protected

area. Such personnel report to the appropriate emergency

response facility for accountability prior to completing

any emergency assignments.

Access to in-plant areas that are contaminated is controlled by barriers, signs, locked doors, or personnel

stationed for that purpose. Emergency monitoring teams

are responsible for determining the need for onsite

access control and establishing the proper method through

discussions with technical support center (TSC)

personnel. Plant procedures used for determining

contaminated areas will be used for determining the need

for access control.

Any food, tobacco, or potable liquids that are inside a radiation or contamination controlled area, regardless of

the packaging, will be considered to be contaminated

until surveyed or otherwise determined to be free of

contamination. These areas will be controlled by plant

health physics procedures and no eating, smoking, or

drinking will be allowed. The emergency director or

designee will make arrangements for supplies to be

brought in.

The emergency health physics supervisor is responsible for permitting return of onsite areas and equipment to normal use once monitoring and decontamination are completed.

REV 19 1/94 TABLE K-1 EMERGENCY EXPOSURE LIMITS FOR WORKERS PERFORMING EMERGENCY SERVICES Dose Limit (rem)(a)

Total Effective Dose Equivalent (TEDE)

Activity Condition 5 All 10 Protecting valuable property.

Lower dose not practicable.

25 Life saving or protection of large populations.

Lower dose not practicable.

>25 Life saving or protection of large populations.

Only on a voluntary basis to persons fully aware of the risks involved.

a. Limit dose to the lens of the eye to three times the listed value and doses to any other organ (including skin and extremities) to ten times the listed value.

L-1 REV 60 07/13 L. MEDICAL AND PUBLIC HEALTH SUPPORT L.1 ONSITE CAPABILITY Provisions have been made to assist personnel who are injured, who may have received high radiation doses, or who have been externally contaminated. Decontamination

materials and portable first aid kits are available at

strategic locations throughout the site. There are

personnel on shift and in the Emergency Organization

trained in first aid and decontamination procedures.

The onsite personnel responsible for responding to a

medical emergency have had training per the OSHA

standard 29 CFR 1910.151 and directive OSHA 3317-06N, Fundamentals of a Workplace First-Aid Program. Health physics technicians will be assigned to first aid teams

in accordance with Procedure 70302-C, Reporting and

Documenting Occupational Injuries Or Illnesses. The

health physics technician will direct and assist in

decontamination of injured personnel as necessary.

Personnel to perform first aid and decontamination will

be available 24-h a day as identified in Section B, Table B-1. In addition, an onsite first aid and decontamination area equipped with decontamination supplies and other

specialized equipment is located near the health physics

station on the 220 ft elevation in the control building.

Personnel found to be externally contaminated but not

requiring immediate medical attention will undergo

decontamination in accordance with plant procedures. L.2 MEDICAL TRANSPORTATION Injured and contaminated personnel requiring hospital medical attention will be transported to Doctors

Hospital, Augusta, or Burke Medical Center, Waynesboro, by the Burke County Ambulance Service.

The ambulance will be met at the plant entry and security building, provided with dosimetry, and

escorted to the patient pickup location. The ambulance

and crew will be prepared to receive the contaminated

patient as time permits by laying protective material

in the ambulance and wearing protective clothing by the

crew. Medical treatment of serious injuries will take

precedence over contamination control. A health

physics technician (HPT) will accompany the patient to

the hospital. The HPT will provide advice and guidance

to the ambulance crew and hospital staff regarding

contamination control, decontamination, radiation

exposure, and protective actions. The HPT will perform

radiation surveys of the patient, ambulance, and

attending hospital staff and assist in maintaining L-2 REV 56 09/10 contamination control in the hospital. All contaminated materials will be properly packaged and

returned to the site for disposal. The ambulance will

also be decontaminated at VEGP if required. L.3 OFFSITE SUPPORT SERVICES Arrangements for treating externally contaminated patients have been made with the Doctors Hospital in

Augusta, Georgia, and Burke Medical Center in Waynesboro, Georgia. The objectives of the support hospital are: (1) to resuscitate and stabilize an

accident victim from a nuclear facility; and (2) to

decontaminate a contaminated/injured patient while

controlling the contaminants to a preselected area

within the hospital. Each hospital has a radiation

emergency area (REA) which has a separate entrance

adjacent to the emergency room complex. This area will

serve as the facility for handling contaminated and

injured patients for the following reasons:

• Its proximity to emergency medical equipment and

supplies.

• Effective isolation of the area from the hospital

proper for contamination control purposes.

• Separate patient ingress and egress for effective

contamination control.

• The area can be denied to routine hospital use for

a given period of time for patient treatment and

decontamination, as well as decontamination of the

area itself.

• Capability to delineate contamination zone (treatment area) from a "clean" area (buffer zone).

• Water supply in treatment area.

Each hospital has available specialized supplies and equipment for decontamination, exposure evaluation, and

contamination control; specifically, a decontamination

table top which provides for washing contamination from

a patient and collecting water in portable storage

containers. Precut plastic material for the floor of

the treatment room and buffer zone are also available to

limit the spread of contamination. Ropes and signs will

be used to delineate the REA. These facilities will

enable emergency treatment and handling of contaminated

individuals. Noncontamination injuries will be handled

by the hospital with its routine facilities.

The medical staff of the hospital is trained to treat externally contaminated patients or individuals who have

received high exposures. Trained plant radiation L-3 REV 62 10/14 protection personnel will assist hospital staff when plant personnel are being evaluated. Following decontamination, personnel suspected to have ingested

radionuclides will undergo whole body counting at VEGP.

In addition, protection of attending personnel is afforded by the availability of disposable clothing and

self-reading and badge-type dosimeters. Radiation

survey instruments are available for monitoring patients

and attending personnel. Each hospital has a

comprehensive kit containing items necessary for removal

of radioactive contamination from skin and wounds.

Included in this kit are culture tubes for nasal, oral, and aural samples, as well as containers for urine, fecal, and wound specimens. Large waste containers and

plastic liners are available for the collection of

contaminated articles.

Each hospital has a procedure entitled, "Decontamination and Treatment of the Radioactively Contaminated

Patient." This protocol details the following

procedures: notification responsibilities of associated

individuals including VEGP health physics personnel, REA

setup, emergency treatment, use of protective clothing, patient decontamination and sample collection

techniques, handling multiple victims, and return of the

REA to routine use. Upon notification of impending

patient arrival, hospital personnel will initiate these

procedures to prepare the REA for patient arrival.

Arrangements have also been made with local doctors to render medical assistance, both onsite and offsite, and

to assume responsibility for the medical supervision of

the patient. These doctors will be on emergency call at

all times and will respond to an accident when called.

Southern Nuclear Operating Company (SNC) or contractor personnel will maintain programs at Doctors Hospital and

Burke Medical Center for treatment of contaminated

individuals. In addition, these personnel provide

training for the hospital staffs. The current contractor

is also on-call for assistance in handling the treatment

of contaminated injured personnel.

In conjunction with the hospital of the University of Alabama in Birmingham, the current contractor has established advanced radiation medical facilities.

These facilities include qualified personnel trained in

the care of radiation accident victims, a radiosurgery

decontamination suite for surgery on contaminated

patients, an exposure treatment suite for treatment of

severely overexposed victims, and an exposure L-4 REV 19 1/94 evaluation laboratory with advanced analysis techniques.

L.4 TRAINING OF MEDICAL SUPPORT PERSONNEL Section O, Radiological Emergency Response Training, identifies the training that will be provided for both

onsite and offsite personnel who have medical support

responsibilities. The VEGP training department conducts

training sessions at least once per calendar year. In

addition, drills and exercises are an integral part of

the training program and are conducted as specified in

section N, Exercises and Drills.

M-1 REV 37 04/03 M. RECOVERY AND REENTRY PLANNING AND POST ACCIDENT OPERATIONS The objectives of Vogtle Electric Generating Plant (VEGP) following any emergency declaration will be to mitigate the

consequences of the event and to take those steps described in

this Emergency Plan which will minimize any effects on the

health and safety of the public and emergency workers. Once the

emergency situation has been terminated, the goal will be to

restore VEGP to normal operating status. For some situations, such as a Notification of Unusual Event involving a natural

phenomenon that has no effect on VEGP, the emergency situation

may not have required any change to normal operations, so no

formal transition will be required. In other circumstances

which may involve suspected or actual damage to the plant, a

transition will be appropriate. This is defined as the recovery

phase. M.1 COMMENCEMENT OF RECOVERY PHASE The emergency director will determine when the recovery phase begins. The following guidelines, as applicable to the specific situation, will be observed prior to

terminating the emergency:

1. The affected reactor is in a stable condition and can be maintained in that condition indefinitely.

2. Plant radiation levels are stable or are decreasing with time.

3. Releases of radioactive material to the environment have ceased or are being controlled within permissible limits.

4. Fire or similar emergency conditions no longer constitute a hazard to safety-related systems or

equipment or personnel.

5. For a site area emergency or general emergency, discussions with plant management, applicable members

of the VEGP emergency organization, offsite authorities

including the Nuclear Regulatory Commission (NRC);

Georgia Emergency Management Agency, Burke County

Emergency Management Agency director; South Carolina

Emergency Management Division director; and Savannah River Site (SRS) emergency staff do not result in

identification of any valid reason for not terminating

the emergency. Once the above conditions have been satisfied, the emergency director will announce that the emergency is

terminated and the plant is in a recovery mode. He will

direct that all elements of the emergency response

organization be advised of the change in status via the

Emergency Notification Network (ENN), Emergency

Notification System (ENS), and other pertinent M-2 REV 19 1/94 communications systems. At this time, the emergency director will designate a recovery manager to constitute

the recovery organization.

M.2 RECOVERY ORGANIZATION Initially, the recovery manager may direct operations from the emergency operations facility (EOF). He will structure the recovery organization to accomplish the

following general objectives:

1. Maintain comprehensive radiation surveillance of the site until levels return to normal.

2. Control access to the affected area of the plant and exposures to workers.

3. Decontaminate affected areas and equipment.

4. Conduct activities in radiation areas in accordance with the plant's standard radiation work practices.

5. Isolate and repair damaged systems.

6. Document proceedings of the accident and review the effectiveness of the emergency response organization in mitigating plant damage and reducing radiation

exposures to the public.

7. Provide offsite authorities with plant status reports and information concerning the plant

recovery organization.

8. Provide assistance with recovery activities undertaken by State and county authorities, if

requested.

9. Provide public information on the status of recovery operations via releases to the media. Individuals will be assigned to specific positions by the recovery manager, depending on the nature and extent

of damage to the plant. Figure M-1 shows a

representative organization for recovery operations.

The responsibilities and functions of the managers shown

on figure M-1 are summarized as follows:

Recovery manager:

has overall responsibility for restoring the station to a normal operating

configuration. Plant operations manager:

manages day-to-day in-plant operations and during recovery, is responsible

for ensuring that repairs and modifications will

optimize post recovery plant operational

effectiveness and safety.

M-3 REV 60 07/13

• Design and construction support manager: focuses necessary engineering, design, and construction resources on those aspects of plant recovery

requiring redesign, modifications, or new

construction; directs and coordinates nuclear steam

supply system (NSSS) and balance-of-plant

engineering and construction/repair work.

• Chemistry manager: develops plans and procedures to process and control liquid, gaseous, and solid

wastes to minimize adverse effects on the health and

safety of the public and plant recovery personnel.

In addition, the chemistry manager coordinates the

activities of staff radiological engineers and

radiation protection personnel engaged in waste

treatment operations.

• Health physics manager: responsible for As Low As Reasonably Achievable (ALARA) planning, execution, and monitoring; plans and manages decontamination of

affected areas and equipment; supervises and directs

all special radiological controls required to

support the recovery operation; and coordinates

environmental assessment activities with the

manager-environmental services.

• Technical support manager: provides analyses, plans, schedules, and procedures in direct support

of plant operations.

• Nuclear oversight manager: ensures that the overall conduct of recovery operations is performed in

accordance with corporate policy and rules and

regulations governing activities which may affect

public health and safety.

• Scheduling/planning manager: prepares plans and schedules and tracks/expedites recovery operations.

• Administration/logistics manager: supplies administrative, logistic, communications, and

personnel support for the recovery operation.

• Public information director: coordinates the flow of media information concerning recovery operations.

Once the organization is established and specific responsibilities are assigned, the recovery manager may

relocate some or all of the recovery organization staff

from the EOF.

M-4 REV 60 07/13 M.3 REENTRY PLANNING If the accident situation involved a release of radioactivity, appropriate areas of the plant and site will

be monitored to determine contamination and radiation levels.

Identifying and controlling access to these areas will be in

accordance with normal plant procedures. When reentry to a

radiation area is required for inspection or work, the

activity will be preplanned and plant radiation work

practices and As Low As Reasonably Achievable (ALARA) program

principles will be followed.

M.4 EXPOSURE MONITORING All personnel who require access to the plant or to radiation areas on site during the recovery phase will be issued

dosimetry, as appropriate. The criteria for reading

Optically Stimulated Luminescent Dosimeters (OSLDs) and assessing radiation dose will be in accordance with standard

health physics practices. The results of the dosimeter

readings, including integrated exposures (i.e., man-rems)

will be reported to the recovery manager, the radcon/radwaste

manager, and others in the plant organization who normally

receive such reports.

The State of Georgia, the State of South Carolina, and SRS

have the responsibility for determining population exposure

of the public via plume exposure and ingestion pathways.

VEGP will provide radiological information including

estimated quantity of radioactivity released, isotopic

composition of released material, and meteorological data to

assist the governmental authorities in their determinations.

By determining the affected population (see appendix 6) and

by performing dose assessment calculations including

determining the quantity of radioactivity and release rate, VEGP personnel can estimate the population exposure if

necessary. Use of data from fixed monitoring stations (OSLDs and air samplers) can be used to confirm the exposure

estimates.

CORPORATE OFFICE NUCLEAR OVERSIGHT MANAGERRECOVERY MANAGER ADMINISTRATIVE/ LOGISTICS MANAGER PUBLIC INFORMATION DIRECTOR SHEDULING/

PLANNING MANAGER TECHNICAL SUPPORT MANAGER PLANT OPERATIONS MANAGER CHEMISTRY MANAGER DESIGN AND CONSTRUCTION SUPPORT MANAGER HEALTH PHYSICS MANAGER Vogtle Electric Generating Plant RECOVERY ORGANIZATION FIGURE M-1 REV 60 7/13 N-1 REV 61 06/14 N. EXERCISES AND DRILLS Vogtle Electric Generating Plant (VEGP) maintains an emergency drill and exercise program in accordance with 10 CFR 50 Appendix E.IV.F. As part of this program emergency exercises and drills are conducted to test and evaluate the adequacy of emergency facilities, equipment, procedures, communication links, actions of emergency response personnel, and coordination between the VEGP and offsite emergency response organizations. Some exercises and/or drills will be

unannounced.

N.1 EXERCISES Emergency preparedness exercises test integrated response capabilities and are conducted in accordance

with Nuclear Regulatory Commission (NRC) and Federal

Emergency Management Agency (FEMA) guidance as described

below. Exercises are conducted every two calendar

years, and are designed to include the demonstration of

response to a major portion of the basic elements of the

emergency preparedness plans of the participating

organizations. The planning and execution of the

emergency exercise is coordinated with Federal, State, and local agencies, as appropriate.

In accordance with 10CFR50, Appendix E, VEGP has not identified any States or local governments refusing to

participate, pursuant to 10 CFR 50.47(c)(1).

For those exercises during which offsite response groups play a significant part, mobilization of Federal, State, and local personnel and resources adequate to verify the

capability to respond to an accident situation is

included. The exercise program for VEGP consists of an 8 year cycle that incorporates the following features:

1. A full participation exercise which tests as much of the Plant Vogtle, State, and local emergency plans as is reasonable achievable without mandatory public participation will be conducted on a biennial basis and evaluated by NRC and FEMA.

2. Biennial exercise scenarios will be submitted to the NRC under § 50.4 at least 60 days before use in the biennial exercise.

3. Each biennial exercise scenario will provide the opportunity for the ERO to perform their key skills as applicable to their emergency response duties in the CR, TSC, OSC, EOF, and joint information center to implement the EP principal functional areas.

N-2 REV 61 06/14

4. Biennial evaluated exercises will be varied such that the following scenario elements are demonstrated over the course of an 8 year exercise cycle:

• Hostile action directed at the plant site.

• No radiological release or an unplanned minimal radiological release that does not require public protective actions.

• Initial classification of or rapid escalation to a Site Area Emergency or General Emergency.

• Implementation of strategies, procedures, and guidance developed under 10 CFR 50.54(hh)(2).

• Integration of offsite resources with onsite response.

5. An ingestion pathway exercise will be conducted on a frequency to ensure the applicable States have the opportunity to participate in an ingestion pathway exercise at least once every exercise cycle. N.2 DRILLS A drill is a supervised instruction period aimed at testing, developing, and maintaining skills in a

particular operation. Drills may be incorporated

into the biennial exercise; they will be supervised

and evaluated by either a training instructor or

designated controller/evaluator.

1. OFF YEAR DRILL Drills shall be conducted to ensure that adequate emergency response capabilities are maintained in

the interval between biennial exercises. At least

one of these drills will be conducted during the

calendar year when there is no biennial exercise and

shall involve a combination of some of the principal

functional areas of the onsite emergency response

capabilities. The principal functional areas include

activities such as event classification, notification of offsite authorities, command and control of emergency response, accident assessment, protective action recommendation development, protective action decision making, and plant system

repair and corrective actions. Activation of all

onsite emergency response facilities (TSC, OSC AND

EOF) is not required during these drills.

Supervised instruction, success paths and accident

management strategies may be included in these

drills.

N-3 REV 61 06/14 The States of Georgia and South Carolina including the Counties of Burke, Aiken, Allendale and Barnwell

will be permitted to participate in drills when

requested by the State or County Government. During each exercise cycle periodic drills will be conducted to ensure the ERO teams (not necessarily each individual)are provided the opportunity to develop and maintain key emergency response skills within the scope of their duties. The ERO (not necessarily each ERO team) will be provided the opportunity to demonstrate key skills in response to the following scenario elements in drills or exercises.

• All functions in each ERF (e.g., all ERFs that are responsible for dose assessment perform those duties in response to a radiological release).

• Use of alternative facilities to stage the ERO for rapid activation during hostile action.

• Real-time staffing of facilities during off-hours (i.e., 6:00 p.m. to 4:00 a.m.).

• Providing medical care for injured, contaminated personnel (every 2 years).

• Response to essentially 100 percent of initiating conditions identified in the site emergency plan implementing procedure for classification of emergencies.

• Response to actual industry event sequences appropriate for the nuclear plant technology (e.g., BWR).

• Use of procedures developed in response to an aircraft threat and in compliance with 10 CFR 50.54(hh)(1).

• Use of the strategies associated with 10 CFR 50.54(hh)(2) to mitigate spent fuel pool damage scenarios (all strategies, such as makeup, spray, and leakage control, but not every variation of a given strategy).

• Use of the strategies associated with 10 CFR 50.54 9(hh)(2) to mitigate reactor accidents and maintain containment).

N-4 REV 61 06/14

2. COMMUNICATION DRILLS To ensure that emergency communication channels between VEGP and offsite authorities are operable, periodic drills are conducted. For drills, the

communication is initiated at VEGP using the actual

message format in accordance with the applicable

plan and procedure. By using the standard message

format, the drill tests understanding of message

content as well as the communication systems

hardware. The following test and drills are

conducted:

a. Communication channels with the State of Georgia, Burke County, State of South Carolina, Aiken

County, Barnwell County, Allendale County, Savannah River Site (SRS), and the NRC are tested

as described in section F.

b. Communication drills among the control room, technical support center (TSC), operations

support center (OSC), emergency operations

facility (EOF), Joint Information Center (JIC),

general office operations center, the States of

Georgia and South Carolina, Burke, Aiken, Barnwell, and Allendale Counties, SRS, and VEGP

field monitoring teams are conducted every two

years. These drills are normally conducted

during the biennial exercise.

3. FIRE DRILLS Fire drills are conducted in accordance with the Final Safety Analysis Report (FSAR). As discussed

in the FSAR, the program involves quarterly drills, at least one of which is unannounced. The quarterly

drills are scheduled so that each member of the fire

brigade participates in at least two drills per

year. In addition, an annual practice is conducted

which requires extinguishing a fire.

4. MEDICAL EMERGENCY DRILLS A medical emergency drill involving a simulated contaminated person is conducted each calendar year.

A moulage kit may be used to mock the injuries;

otherwise, the drill script will describe the

injuries. Controller data will be used for dose

rates, contamination levels, and dosimeter readings.

The simulated injured player is given initial

treatment, as appropriate, by the VEGP first aid

team, transported by ambulance to the hospital, and

then given subsequent treatment by the hospital

staff. This drill may be conducted as part of the

biennial exercise.

N-5 REV 61 06/14

5. RADIOLOGICAL MONITORING DRILLS Plant environs and radiological monitoring drills are conducted each calendar year per Procedure NMP-

EP-303, Drill and Exercise Standards. For these

drills, a team is dispatched with a controller to

obtain the required measurements or samples. The

drill controller will evaluate the proper use of

survey instruments, record keeping, communications, and the collection of sample media (soil, air, water, and vegetation), as appropriate. Procedures

to be tested include Procedure 91303-C. This drill

may be conducted in conjunction with one of the

semiannual health physics drills.

6. HEALTH PHYSICS DRILLS Semiannual health physics drills will be conducted per Procedure NMP-EP-303, Drill and Exercise

Standards. The emergency preparedness supervisor

will direct the development of drill scenarios so as

to test and evaluate the response activities. Drill

scenarios will simulate, as closely as possible, actual anticipated emergency conditions. Simulated

elevated airborne and liquid samples and radiation

in the environment will be utilized for drill

activities. Drills will evaluate the proper response per Procedure 91302-C and Procedure 91303-C. Use of

sample techniques, survey techniques, monitoring

methods, decontamination methods, and protective

clothing and respirators will be demonstrated as

appropriate during the drill but may not be used

throughout the drill (for example, field monitoring

teams will not wear protective clothing). Exposure

control considerations will also be exercised during

the drill. Post accident sampling under simulated accident conditions will be demonstrated each calendar year.

The post accident analysis may be performed using

available instrumentation or using laboratory

equipment to demonstrate the methods employed under

actual accident conditions.

N.3 SCENARIOS Each drill and exercise is conducted in accordance with a scenario. The scenarios for the drills may be

considerably less extensive than the scenario for the

biennial exercise. The preparation of exercise

scenarios is directed by the emergency preparedness

supervisor who enlists the assistance of personnel from

other departments, as required, to assist in this task.

N-6 REV 61 06/14 The scenario is coordinated with offsite authorities when they are participating in the exercise. The

scenario is submitted to the NRC in accordance with

current practice. FEMA receives a copy of the scenario

from the participating states.

Scenarios include the following information:

• Objectives.

• Date, time period, place, and participating organizations.

• Controller/evaluator assignments.

• Time schedules of real and simulated initiating events.

• Messages describing equipment malfunctions, personnel injuries, and other non plant events, as

appropriate.

• Narrative summary describing the conduct of the drill or exercise.

• Radiological data for onsite facilities and offsite field monitoring teams. The exercise program is structured with sufficient flexibility to allow free play for decision making

processes. The exercise scenario package identifies a

specific accident sequence and includes messages that

support the accident sequence. The exercise control

organization receives general instructions concerning

the deviation of plant personnel from procedural

response. The exercise control organization may

restrict player action if the response would interfere

with the time sequence or prevent demonstration of an

exercise objective. Free play items may be included in

the scenario to maintain player interest.

Specific elements that may allow free play in the decision making process during the exercise include:

• Damage control.

• Accident mitigation.

• Manpower augmentation actions.

• Exposure control actions.

• Communication with offsite authorities.

• Recommendation of protective actions.

N-7 REV 61 06/14 N.4 EVALUATIONS AND CORRECTIVE ACTIONS All drills and exercises are evaluated. For periodic drills, the process consists of the following steps:

1. Drills will be evaluated by controllers/evaluators selected on the basis of expertise and availability.

2. Improper or incorrect performance during the drill may be corrected by the controller/evaluator and the

proper method pointed out or demonstrated.

3. The exercise or drill controllers assemble the players at the conclusion of activities for a

critique. Players are encouraged to identify areas

where improvements are required. The drill

controllers also present their observations to the

players.

4. The site emergency preparedness supervisor submits a list of corrective actions, responsibilities, and

schedule information to the site support manager for

approval.

5. The emergency preparedness supervisor assigns action items and monitors the status of completion of

corrective actions. Significant problems will be

brought to the attention of appropriate plant

management. Exercise evaluation and corrective action are carried out in similar fashion. Critiques with the players are

conducted in each facility and coordinated by the

controller/evaluator at that facility. Each

controller/evaluator submits written reports to the

exercise controller. An overall critique is presented

to key players and to the control organization after the

exercise. The site support manager approves the

responsibilities for corrective actions and deadlines

for completion. The site emergency preparedness

supervisor monitors completion status.

In addition to the internal critique and evaluation, Federal observers may observe, evaluate, and critique

the biennial exercise. Corrective actions resulting

from this critique, together with deadlines for

completion, are assigned by the site support manager.

The site support manager is advised of the status of

these open items periodically.

As stated in 10CFR50, Appendix E, section IV, if VEGP failed to demonstrate with reasonable assurance that

protective measures can and will be taken, a remedial

exercise would be performed as directed by the NRC.

O-1 REV 35 6/02 O. RADIOLOGICAL EMERGENCY RESPONSE TRAINING Emergency response training is provided at four levels:

1. All VEGP badged personnel will receive General Employee Training at inception of onsite duties.

GET will include emergency classification, individual response, signals, accountability, and

site dismissal procedures.

2. All VEGP emergency response organization personnel will receive specialized training per Table O-2.

3. Offsite response groups who may support onsite situations, such as fire or personnel injury, will

be offered annual training in notification, expected

roles, site orientation, security procedures, and

basic radiation protection.

4. Selected state and local emergency response management personnel with offsite emergency response

roles will be offered a seminar/training course in

specific areas; (1) the VEGP emergency

classification system; (2) the VEGP protective

action recommendation criteria and their

relationship to plant conditions; and (3) the VEGP

emergency response organization. These offsite

management personnel will be offered initial

training and annual retraining. Coordination with

offsite authorities will include planning for and

participation in VEGP exercises.

O.1 TRAINING As a minimum, training will be provided in the subject areas shown in table O-1 to various personnel according

to their emergency response position as shown on table

O-2. It should be noted that these subject areas do not

necessarily represent specific course titles, since

several individual courses may be used to implement the

training in each area. Also, both the content and depth

of training may be varied slightly, depending upon the

particular audience, in order to tailor the presentation

to the specific needs of the group.

The training will be conducted in accordance with lesson plans. Classroom lectures, demonstration and use of

equipment, and walk throughs of facilities will be

incorporated into the lesson plans as appropriate. A

written examination or practical exercise may be

administered at the conclusion of a lesson. Records of

the attendance and examination scores will be retained

in the training files. Those designated to receive

training in each subject area are indicated in

table O-2.

O-2 REV 50 11/08 Radiological emergency response training is offered throughout the year, with each training course being covered at least once per calendar year or as often as

necessary to ensure that ERO personnel remain qualified

per training requirements in section O.2. Annual

retraining consists of initial training material

reinforcement and appropriate lessons learned from the

previous year's operating experience. Lessons learned

that are distributed by other methods may not be

included in annual retraining. The plant manager may receive credit for Management of Radiological

Emergencies (MRE) requalification by participating in an

integrated drill or annual exercise.

In addition, drills and exercises are an integral part of the training program and are conducted as specified

in section N of this Plan. During practical drills, on

the spot corrections will be made, if the situation and

time allow. If not, the corrections will be pointed out

in the critique. Upon completion of each training

session or drill, the participants will be asked to

critique the training in order to ensure continued

improvement.

O.2 QUALIFICATION Emergency response personnel at VEGP are qualified by the following criteria:

Normal duties Positions in the emergency organization are assigned commensurate with normal

managerial, supervisory, and/or

technical skills, as shown in

table B-2.

Mandatory training for

emergency positions held All ERO personnel shall be

trained per table O-2 within

the last 15 months, except for

post accident sampling and

first aid training, which is to

be within 36 months.

Drills/exercise performance Individual performance is evaluated and corrective

actions taken as necessary.

O.3

SUMMARY

All badged VEGP workers will receive general training in emergency preparedness. Selected individuals on site and off site will annually receive specialized training

in order to implement the VEGP Emergency Plan.

REV 57 06/12 TABLE O-1 (SHEET 1 OF 3)

TRAINING COURSE DESCRIPTIONS Training Course Description Core Damage Assessment This course covers the calculational methodology for assessing core damage and estimating potential

source terms. It includes retrieval

of pertinent plant parameter data

from the control room; making core

inventory determinations based on

reactor power history; estimating

cladding and/or fuel damage; and

estimating resultant activity

released to containment atmosphere.

Offsite Communications This course covers operation of communications equipment in the

ERFs; communications methods, and

procedures for notification of off-

site emergency response agencies.

Emergency Plan Overview(a)

(EPO) This course covers an overview of

the Emergency Plan with special

attention to emergency planning

zones (EPZs); emergency

classification system; onsite

emergency response organizations;

responsibilities of emergency

response personnel; and site

accountability and site dismissal.

First Aid This course covers the required subject material as referenced in OSHA 3317-06N 2006, "Best Practices Guide: Fundamentals of a Workplace First Aid Program."

a. EPO is included in G.E.T. badge training for all unescorted personnel.

REV 57 06/12 TABLE O-1 (SHEET 2 OF 3)

Training Course Description Management of Radiological

Emergencies (MRE)

This course covers classification of

emergencies; emergency notification

of on and offsite emergency response

personnel and agencies; activation

and staffing of emergency response

facilities; protective action

recommendation decision-making based

on EPA PAG; retrieval of available

plant computer data; reentry and

repair operations; and

communications and information

management; and recovery.

Offsite Dose Assessment This course covers dose projection methodology including computerized

methods; retrieval of plant computer

data; methods for obtaining

meteorological data; operation of

the dose assessment computer; and

interpretation of offsite dose

calculation results.

Post-Accident Sampling (TSC Chemistry Supervisors

Only) This course covers collection of

samples from plant process and

effluent streams under emergency

conditions; measuring radionuclide

and selected chemical concentrations

in those samples; interpretation of

sample results. Training requirement

for the TSC chemistry supervisor is

triennial.

Repair and Corrective

Actions This course covers As Low As

Reasonably Achievable (ALARA)

principles as they apply to planning

and implementing repair and

corrective action; emergency

exposure guidelines; and

communications during repair and

corrective actions.

Field Monitoring Team This course covers field measurement of airborne radioactivity, radiation

levels and contamination in the EPZ;

collecting environmental samples;

map reading; record keeping; and

radio communications.

REV 39 12/04 TABLE O-1 (SHEET 3 OF 3)

Training Course Description Radiological Emergency

Team In Plant This course covers methods for

performing in-plant radiation, contamination, and airborne

radioactivity surveys under

emergency conditions; search and

rescue of missing personnel;

managing health physics activities; and communications for the above

activities.

Security This course covers emergency response activities of the security

department including personnel

accountability; traffic control;

communications; and access control

to emergency response facilities.

Medical Support of

Radiation Emergencies This course covers the

responsibilities and methods for

handling exposed and/or contaminated

injuries. It includes interfacing

with ambulance crews; health physics

activities for transporting a

contaminated injured patient to the

hospital; monitoring and

decontamination while at the

hospital; and final disposition of

the ambulance, ambulance crew, hospital staff, and radiation

emergency area (REA) at the

hospital.

SCBA This course covers the use of self-contained breathing apparatus

including equipment description;

proper donning and use; and

inspection and actions in case of

equipment failure.

REV 56 09/10 TABLE O-2 (Sheet 1 of 2)

Training Requirements For VEGP ERO Personnel CORE DAMAGE ASSESSMENT OFFSITE COMMUNICATIONS EMERGENCY PLAN OVERVIEW FIRST AID MANAGEMENT OF RADIOLOGICAL EMERGENCIES OFFSITE DOSE ASSESSMENT POST-ACCIDENT SAMPLING REPAIR AND CORRECTIVE ACTIONS FIELD MONITORING TEAM RAD EMERGENCY TEAM IN-PLANT

SECURITY MEDICAL SUPPORT OF RADIOLOGICAL EMERGENCY SCBA Emergency Director x x X EOF Management - Training provided as described in Appendix 7 EOF Staff

- Training provided as described in Appendix 7

Dose Analyst x x Security Coordinators x x TSC Manager x x TSC Support Coordinator x Engineering Supervisor x x Maintenance Supervisor x x Operations Supervisor x x Health Physics Supervisor x x x x Chemistry Supervisor x x TSC Engineering Staff x OSC Manager x x x Communicators x x Clerks x Teams In-Plant Monitoring x x x Damage Control/Assessment x x x Repair And Modification x x x Search And Rescue x X(a) x Fire Brigade x x First Aid x x x REV 46 07/07 TABLE O-2 (Sheet 2 of 2)

Training Requirements For VEGP ERO Personnel CORE DAMAGE ASSESSMENT OFFSITE COMMUNICATIONS EMERGENCY PLAN OVERVIEW FIRST AID MANAGEMENT OF RADIOLOGICAL EMERGENCIES OFFSITE DOSE ASSESSMENT POST-ACCIDENT SAMPLING

REPAIR AND CORRECTIVE ACTIONS FIELD MONITORING TEAM RAD EMERGENCY TEAM IN-PLANT

SECURITY MEDICAL SUPPORT OF RADIOLOGICAL EMERGENCY SCBA Field Monitoring Team x x Dosimetry x Health Physics Technicians x x x x Monitoring Team Communicator x x

(a) At least one member first aid qualified

P-1 REV 62 10/14 P. RESPONSIBILITY FOR THE PLANNING EFFORT The Executive Vice President/Chief Nuclear Officer (CNO)

Southern Nuclear Operating Company (SNC) has overall

responsibility and authority for all nuclear activities, including emergency planning (EP) programs. Reporting to the

Executive Vice President of Operational Readiness and Site Integration is the Site Integration Director-(Plant).

The SNC Emergency Planning program is comprised of two

distinct and integral functions; emergency planning and

emergency preparedness. Responsibility for the performance

of these functions is assigned to various members of the SNC

Organization and coordinated as follows.

Emergency Planning:

The Vice President Regulatory Affairs reports to the president/CEO. This individual is responsible for licensing

through: providing organizational support and management

oversight of the sites to assure prompt and proper

disposition of regulatory issues; the development of

regulatory positions; advising senior management on

priorities and activities affecting regulatory at the nuclear

sites; and interfacing with NRC management on behalf of the

sites. Other responsibilities include: developing policies, standardized processes, and procedures for the maintenance of

the licensing basis; the preparation of submittals to the NRC

and other regulatory organizations; and the dissemination of

regulatory information. Reporting to the vice president-

regulatory affairs is the fleet emergency preparedness

manager, the fleet performance improvement manager, the

regulatory affairs director-fleet, and the regulatory affairs

director-nuclear development. The regulatory affairs

director-nuclear development is functionally independent of

SNC's operating fleet and is noted here for completeness

only. Accordingly, the vice president-regulatory affairs is

responsible for administration of the corrective action

program in the corporate headquarters, the overall

coordination of the corporate emergency preparedness programs (including the common Emergency Operations Facility),

Emergency Plans, and site emergency response communication.

His direct report, the Fleet Emergency Preparedness Manager, has the overall governance, oversight, and support of fleet

emergency preparedness activities and programs.

The Fleet Emergency Preparedness Manager is responsible for

overseeing emergency planning activities offsite and

coordinating those activities with Licensee, Federal, State

and local response organizations. The Emergency Planning

Coordinator(s) reports to the Fleet Emergency Preparedness

Manager in support of this effort.

P-2 REV 62 10/14 The Emergency Plans are maintained by the Fleet Emergency Preparedness Manager. The Fleet Emergency Preparedness

Manager provides strategic direction for SNC emergency

planning and coordinates with site management through the

Vice President - Fleet Operations Support.

The Emergency Planning Coordinator(s) coordinate site

input and involvement in emergency planning programs with

the Emergency Preparedness Supervisor. The Emergency

Planning Coordinator(s) review Emergency Plan changes to

determine if the effectiveness of the specific plans have

been reduced. Emergency Plan changes which are judged to

reduce the effectiveness of the Plan will be submitted to

the NRC for approval prior to implementation.

Emergency Preparedness:

The Vice President - (Plant) is responsible for the site Emergency Preparedness aspects of the program. The Emergency

Preparedness Supervisor is responsible for coordinating

emergency preparedness activities onsite and supports offsite

emergency preparedness activities in the vicinity of the

plant. The Emergency Preparedness Supervisor reports through

the Site Integration Director to the Executive Vice President of Operational Readiness and Site Integration-(Plant).

The Emergency Planning Supervisor is responsible to the Site Integration Director for implementation of emergency planning strategies.

Coordination

The Fleet Emergency Preparedness Manager coordinates site

input and involvement in emergency planning programs with the

Emergency Preparedness Supervisor. The Emergency Preparedness

Supervisor is responsible for the implementation of the

Emergency Plan and procedure development and maintenance.

Figure P-1 shows the EP Organization. The Fleet Emergency

Preparedness Manager, Emergency Planning Coordinator, Emergency Preparedness Supervisor, and other individuals with

emergency planning responsibilities are trained by self-study

and by attending industry seminars, short courses, workshops, etc.

Onsite Emergency Plan Implementing Procedures (EPIP) are

maintained by the Regulatory Affairs Manager with the

Emergency Preparedness Supervisor being the principal site

contact. EPIPs for the corporate emergency response

organization and procedures governing fleet emergency

planning activities are maintained by the Fleet Emergency

Preparedness Manager. The Fleet Emergency Preparedness

Manager performs a review of the site specific emergency plan

annually and all onsite EPIPs biennially. The review P-3 REV 61 06/14 includes the letters of agreement, which are updated as necessary.

The Fleet Emergency Preparedness Manager performs a review of

the emergency plans for Southern Nuclear once each calendar

year. The review includes a comparison for consistency of

all emergency plans for the specific sites including the

Security Plan, State, County, and the Savannah River Site

plan as appropriate.

The Emergency plans and EPIPs are revised in accordance with applicable site procedures.

Emergency Plan changes which are judged to reduce the effectiveness of the Plan will be submitted to the NRC for

approval prior to implementation. The Emergency Planning

Coordinator will review Emergency Plan changes to determine

if the effectiveness of the site specific plan has been

reduced prior to submitting the proposed change for

departmental review and subsequently to the PRB for approval.

As required by 10 CFR 50.54(t), an annual independent audit

of the emergency preparedness program is conducted by the SNC

Nuclear Oversight Department. This audit is conducted as part of the standard audit program and will include a review

of the Emergency Plan, its implementing procedures and

practices, emergency preparedness training, annual exercises, equipment, and emergency response facilities. In addition, an audit of the interfaces with offsite agencies is performed

by the SNC Nuclear Oversight Department.

Each audit is nominally conducted every 12 months.

Audits are performed in accordance with SNC Nuclear Oversight department procedures. Audit reports are written and

distributed to management and, in addition, applicable

portions of the corporate audit reports are made available to

affected Federal, State, and local agencies, as appropriate, in accordance with 10 CFR 50.54(t).

Appropriate departments are responsible for implementing corrective actions resulting from the audit findings.

Records of these audits and exercise findings are maintained in accordance with plant procedures.

In addition to this Plan, several other formal emergency

plans have been developed to support the overall

emergency response effort. These supporting plans and

their sources are listed in procedure NMP-EP-300, SNC

Corporate Emergency Planning Activities.

P-4 REV 62 10/14 Chairman, President, & CEOEVP Operational Readiness & Site IntegrationOperational Readiness Vice PresidentSite Integration DirectorVogtle 3 & 4ProcedureVogtle 3 & 4 Security/

EP Project ManagerFleet Emergency Preparedness ManagerVogtle 1 & 2 SecurityVogtle 1 & 2 EPVogtle 3 & 4 SecurityVogtle 3 & 4 EP FIGURE P-1 EMERGENCY PREPAREDNESS ORGANIZATIONS