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{{Adams
#REDIRECT [[NL-15-1569, Revision 65 to Emergency Plan]]
| number = ML15233A290
| issue date = 08/17/2015
| title = Revision 65 to Emergency Plan
| author name =
| author affiliation = Southern Nuclear Operating Co, Inc
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000424, 05000425
| license number =
| contact person =
| case reference number = NL-15-1569
| package number = ML15233A287
| document type = Emergency Preparedness-Emergency Plan
| page count = 298
}}
 
=Text=
{{#Wiki_filter:REV 65  8/15
 
VOGTLE ELECTRIC GENERATING PLANT UNIT 1 AND UNIT 2 EMERGENCY PLAN
 
JUSTIFICATIONSFORCHANGESREVISION65ChangeJustificationPageH 9:AddedRelativeHumidityasaparametertransmittedtotheControlRoom.DCPSNC143331replacestheDewpointTemperaturesensorwithaRelativeHumiditysensor.ThenewmeteorologicaldataloggercalculatesDewpointTemperatureusingtheRelativeHumidityvalueandAtmosphericTemperaturevalve.TheDewpointtemperatureandRelativeHumidityvalueswillbedisplayedintheControlRoomsvi aIPCdisplays.PageH 13:Themethodofhowmeteorologicalinformationistransmittedtotheplantwasdeleted.Microwavetransmissionofdataisnolongerperformed.METdataissenttotheplantusingfiberopticcables.TableH 1:Deletedreferencetotheprimaryandbackuptower.Revisedtoshowprimaryandbackupsensorsareavailableforwindspeed,winddirection,indicationofstabilityclass,andtemperature.DeletedreferencetoUnit1IPConly.Onlyonetowerwillnowbeused.PrimarydatawillbesenttotheUnit1IPCandsecondarydatawillbesenttotheUnit2IPC.PageI 4:Revisedfourthparagraphandremovedreferencetothebackuptower.Toidentifythattherearefullyredundantsecondary(backup)instrumentsavailableonthe60 mtowerfordispersioncalculations.IndexofEffectivePages ,pagesxiiiandxiv:UpdatedRevisionNumber.Updatedtoreflectcurrentversionnumber.
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 65 8/15 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 65 H-10 35 H-11 35 H-12 56 H-13 65 H-14 38 H-15 38 H-16 22 H-17 34 H-18 57 Table H-1 65 Figure H-1 34 Figure H-2 57 I-1 56 I-2 57 I-3 34 I-4 65 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 VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS PAGE REV. xiv REV 65  8/15 J-6 11 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 63 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 hours (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 hours 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 hours 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 hour 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 hours
{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 hour 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 hours 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 hours 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
 
#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 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 hours 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 hours 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 hours 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 65  8/15 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). Relative humidity (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 65  8/15 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 hours 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 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 65  8/15 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 wind speed, wind direction, indication of stability class, temperature)
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 65  8/15 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
 
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
 
meteorological tower is equipped with secondary backup instruments at the 10-m and 60-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 hours 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}}

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