NL-16-1409, Submittal of Report of Changes to Emergency Plan

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Submittal of Report of Changes to Emergency Plan
ML16291A509
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 08/22/2016
From: Wheat J
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-1409
Download: ML16291A509 (17)
v • d • e


Text

Justin T. Wheat Southern Nuclear Nuclear Licensing Manager Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242

,~ Southern Nuclear Tel 205.992.5998 Fax 205.992.7601 AUG 2 2 2016 Docket Nos.: 50-424' 10 CFR 50.54(q) 50-425 NL-16-1409 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 1 and 2; Report of Changes to Emergency Plan Ladies and Gentlemen:

In accordance with 10 CFR 50.54(q)(5) and 10 CFR 72.44(f), Southern Nuclear Operating Company (SNC) hereby submits descriptions of changes to plant emergency plans and a summary of the analysis demonstrating that the changes did not reduce the effectiveness of the plans. The plans, as changed, continue to meet the requirements in 1o CFR 50 Appendix E and .the planning standards of 10CFR 50.47(1::>).

Description of Changes and Summary of Analysis Effective July 26, 2016, Vogtle Ele_ctric Generating Plant (Vogtle) implemented changes to the emergency preparedness procedure NMP-EP-110-GL03, "VEGP EALs-ICs, Threshold Values arid Basis" and changes to the Vogtle Emergency Plan (Version 67). Several changes resulted from a Vogtle engineering analysis of setpoints for Emergency Operating Procedures (EOPs) and Abnormal Operating Procedures (AOPs).

This analysis refined the setpoints to improve the.accuracy of the values. Specifically, EAL values were changed to align with tolerance improvements described in EOP and AOP setpoint documents. In addition, a reqliireinentfor the TSC Director to have Self .

Contained Breathing Apparatus (SCBA} training was eliminated because the requirement was not related to any emergency planning commitment. Finally, conforming changes were made due to errors* found in implementation procedures that were not in alignment with the approved NEI 99-01 Rev. 4 EAL scheme.

Following the guidance in NRC A.G. 1.219, SNC concluded that the changes to EAL VEllues reflected the design documents a.nd that the meaning and intent of the basis of the approved EAL was unchanged. All of the changes were evaluated in accordance with 10 CFR 50.54{q}(3}; and it was determined that these changes did not reduce the effectiveness of the Vogtle Emergency Plan and the revised procedure and Vogtle Emergency Plan continue to meet the requirements in 1o CFR 50 Appendix E and the planning standards .of 10 CFl150,47(b}.

U.S. Nuclear Regulatory Commission NL-16-1409 Page2 This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) .992-7369.

Respectfully submitted,

  • c;:§)-:C:/-/~ *.

Justin T. Wheat Nuclear Licensing Manager jtwlefb/lac

. ' ~

cc: Southern Nuclear Operating Company

  • Mr. S. E. Kuczynski, Chairman, President & CEO

.Mr. .D. G. Bos( Executive Vice President & Chief Nuclear Officer Mr~ B. K. Taber, Vice President-Vogtle Unit t & 2 Mr. M. D. Meier, Vice President- Regulatory Affairs Mr. ,A. L. Mansfield - Director, Emergency Preparedness Mr. G" W. Gunn, Regulatory Affairs Manager ~Vogtle 1 & 2 RType: Vbgtle=CVC7000 .

iu. s. Nuclear Regulatory Commission .

M~. C. *Haney; Regional Administr,ator Mr; 'A. E. Martin, NRA Senior Proj~ct Manager - Vogtle Ms. N. R. Childs, Resident Inspector - Vogtle State of Georgia

  • Mr. J. H. Turner, Director- Environmental Protection Division
  • VOGTLE ELECTRIC GENERATING PLANT UNIT 1 AND UNIT 2 EMERGENCY PLAN Revision 67 June 2016 Revision Insertion Instructions Please replace the affected pages in your copy of the Plan with the corresponding Revision 67 pages. Pages included in this package are:

Title Page List of Effective Pages page x, xi, xii, xiii, xiv, and xv Section D pages D-13, 0-60, 0-61, 0-62, 0-110, and 0-115 Section 0 Table 0-2 sheet 1 Discard these instructions after use and sign and return Transmittal Acknowledgment to address indicated .

VOGTLE ELECTRIC GENERATING PLANT UNIT 1 AND UNIT 2 EMERGENCY PLAN

  • REV 67 6/16

VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS Table of Contents (i) 51 ii 43 iii 57 iv 21 v 46 vi 27 vii 63 viii 38 ix 63 x 63 xi 67 xii 67 xiii 67 xiv 67 xv 67 xvi 60 xvii 60 xviii 60 Preface xix 39 xx 41 Figure i 5 Figure ii 27

  • Figure iii Figure iv Figure v Figure vi A-1 A-2 14 23 NONE 12 56 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 x REV 67 6/16

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

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 C-2 C-3 C-4 C-5 Table C-1 (1 of 2)

Table C-1 (2 of 2) 61 61 61 61 61 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 67 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 62 D-32 57 D-33 48 Xl REV 67 6/16

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

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 67 D-61 67

  • D-62 D-63 D-64 D-65 D-66 D-67 D-68 to D-71 67 62 53 48 62 62 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 48 D-103 48 D-104 48 D-105 48 D-106 48 D-107 48 D-108 48 D-109 48 XU REV 67 6/16

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

D-110 67 D-111 48 D-112 48 D-113 48 D-114 48 D-115 67 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 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 F-5 F-6 F-7 Table F-1 G-1 G-2 28 57 60 43 60 24 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 xm REV 67 6/16

VEGP EMERGENCY PLAN LIST OF EFFECTIVE PAGES WITH REVISIONS 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 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)

Table J-5 (2 of 2)

Figure J-1 Figure J-2 K-1 K-2 K-3 39 41 63 63 57 21 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 61 N-7 61 0-1 35 0-2 50 Table 0-1 (1 of 3) 57 XIV REV 67 6/16

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

Table 0-1 (2 of 3) 57 Table 0-1 (3 of 3) 39 Table 0-2 (1 of 2) 67 Table 0-2 (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 Appendix 3 0 3-1 61 3-2 19 3-3 5 3-4 56

  • 3-5 3-6 3-7 3-8 3-9 3-10 Figure 3-1 52 31 31 38 38 38 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 61 6-2 61 6-3 61

  • xv REV 67 6/16
  • challenges as YELLOW, ORANGE, and RED paths. If the core exit thermocouples exceed 1200 degrees F or 700 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 Eme-rgency 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 ~

Threshold Value (TV) for an EAL considering each Unit independently. Two IC Matrices are used, one for Hot res and one for Cold res. 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 res 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 a:re six EALs which specifically state that if the condition cannot be mitigated and is imminent, the Emergency D-13 REV 67 6/16

Voatle Fission Product Barrier Evaluation General Emer11encv Site Area Emer11encv Alert Unusual Event FG1 FS1 FA1 FU1 Loss of ANY Two Barriers AND Loss Loss or Potential Loss of ANY ANY Loss or ANY Potential Loss of ANY Loss or ANY Potential Loss of or Potential Loss of Th ird Barrier Two Barriers EITHER Fuel Clad OR RCS Containment Fuel Clad Barrier Potential Loss

1. Critical Safe!J! Function Status 1. Critical Safe!J! Function Status Core-Cooling RED Core Cooling-ORANG E OR Heat Sink-RED
2. Prima!:J! Coolant Activi!J! Level z. ~rima!Jl ~11nt ~l1i!l! l.!!l1!11 Indications of RCS Coolant Activity greater than 300 µCi/gm Dose Not Applicable Eauivalent 1- 131
3. Core Exit Thermocou11le Readings 3. Core Exit Thermocou11le Readings Core Ex it TCs qreater than 1200°F Core Exit TCs qreater than 700°F

~ BH!i!m :llmm Not ADOlicable w.mc L.llt'.111 4. Reactor Vessel Water Level RVLIS LEVEL less than 63%

5. Containment Radiation Monitoring lj. Con!iln!!l!!nt R!!!li!!liQn M!!!!itQring Containment Rad iation Monitor RE-005 OR 006 2 2 6E+5 mR/hr Not Anoticable 6 Other lndicall<>O* !i Qlbm: l!!!licali2a1 Not applicable Not aDDllcable
7. Emergencl( Director Judgment 7. Emergenclr'. Director Judgment Judgment by the ED that the Fuel Clad Barrier is lost. Consider Judgment by the ED that the Fuel Clad Barrier is potentially lost. Consider conditions not addressed and inabil ity to determine the status of the Fuel conditions not addressed and inability to determine th e status of the Fuel Clad Barrier Clad Barrier.

RCS Barrier Potential Loss

1. !<ci!is<l!I Sat8lv E!!as;!!Qn l2l!!!!!I 1. Critical Safe!J! Function Status Not Applicable RCS Integrity-RED OR Heat Sink-RED
2. RCS Leak Rate 2. RCS Leak Rate RCS subcool ing less than 22°F {less than 22° F Adverse} due to an RCS Non-i solable RCS leak (includ ing SG tube Leakage) greater than 120 gpm leak qreater than Cha rq inq I RHR capacity
3. SG Tube Ru11ture a l2~ !111!!! B!!r&IC!!

SGTR resultina in an SI actuation Not Anriticable

4. Containment Radiation Monitoring ~ - !<2DllliD!!l!!DI Bil!li!!li2!l M2!!il2!'.ing CTMT Rad Monitor RE-005 OR 006 2 8.7E+2 mR/hr Not ADDlicable
5. Other lndlcatiQns 5. Other Indications Not applicable Unexplained level rise in ANY of the following :

Containment sump Reactor Coolant Drain Tank (RCDT)

Waste Holdup Tank (WHT)

6. Emergenclr'. Director Judgment 6. Emergenclr'. Director Judgment Judg ment by the ED that the RCS Barri er is lost. Consider conditions not Judgment by the ED that the RCS Barrier is potentially lost. Consider addressed and inability to determine the status of the RCS Barrier cond itions not addressed and inability to determine the status of the RCS Barrier.

Containment Barrier Potential Loss

1. Critical Safe!lr'. Fungjon Status 1. Critical Safe!J! Function Status Not Annlicable Containment-RED
2. Containment Pressure 2. Containment Pressure Rapid unexplained CTMT pressure lowering following initial pressure rise CTMT pressure greater than 52 psig OR OR lntersystem LOCA indicated by CTMT pressure or sump level response CTMT hydrogen concentration greater than 6%

not consisten t with a loss of primary or secondary coolant 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 sprav

a. Q2r11 ~ii It!!!nm!!!<!!1111!!! B!!!!!ling 3. Core Exit Thermocou11le Reading Not applicable CORE COOLING CSF - RED for greater than 15min OR CORE COOLING CSF - ORANGE for greater than 15min AND RVLIS LEVEL less than 63%
4. SG Seconda!:J! Side Release with Prima!:J! to Seconda!:J! Leakage 4. li!~ li!~nd!!!Jt li!i!l!! R!!!!!a§!! with P-tQ:§ L11ak!!ge RUPTURED S/G is also FAUL TED outside of contain ment Not applicable OR Primary-to-Secondary leakrate greater than 10 gpm with nonisolable steam release from affected S/G to the environment
5. CNMT Isolation Valves Status After CNMT Isolation :i. !<t:lMI 112.!i!li!!!l :llilllr'.!ll §lilll!I t.l!!!r QNMI l!!Q!!!!iQ!l CTMT isolation valve(s) OR damper(s) are NOT closed res ulting in a Not Applicable direct pathwav to the environment after con tainment isolation is reau ired

!i l2i9ai!ican! R!!!li!!!!!<lil1!! l!ll1!1n!Q!:J! io !<Qn!S!i!J!!l!!n! 6. Sign ificant Radioactive lnvento!:J! in Containment Not Applicable CTMT Rad monitor RE -005 OR 006 2 1.3E+7 mR/hr

7. Other Indications 7. Oth!!r lndicati2n1 Pathway to the environment exists based on VALID Not applicable RE-2562C Alarm AND RE-12444C OR RE -12442C Alarms
8. Emergenclr'. Director Judgment 8. EmergenCl( Director Judgment Judgment by the ED that the CTMT Barrier is lost. Consider conditions Judgment by th e ED that the CTMT Barrier is potentially lost. Consider not addressed and inability to determine the status of the CTMT Barrier cond itions not addressed and inability to determine the status of the CTMT Barrier D- 6 0 REV 67 6/ 16
  • 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 535 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 indicates 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 700 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-61 REV 67 6/16
  • 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 1-131 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 535 gpm due to.

operator action.

This Threshold Vallie 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-62 REV 67 6/16
  • SYSTEM MALFUNCTION Initiating Condition -- SITE AREA EMERGENCY 554 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 535 gpm due to operator action.

1. Complete Loss of Heat Removal Capability as indicated by:
  • a.

b.

Core Cooling CSF - ORANGE 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 I Radiological Effluent, Emergency Director Judgment, or Fission Product Barrier Degradation ICs .

  • D-110 REV 67 6/16
  • SYSTEM MALFUNCTION Initiating Condition** GENERAL EMERGENCY SG2 Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip was NOT Successful AND there is Indication of an Extreme 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 535 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 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-115 REV 67 6/16

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u Emergency Director 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 Radiation Protection x x x x Supervisor 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 67 6/16

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